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Sample records for acr-1000 fuel bundle

  1. 2009 creep correlations of Zircaloy for ACR-1000® fuel

    The experimental creep data of Zircaloy-4 fuel cladding obtained in the late 1970's has been the main input data for developing creep correlations in CANDU® fuel design. However, these data were measured for those creep strains with short time durations and under a single neutron flux rate. The ACR-1000® fuel bundle has been designed to operate at high burnup, high coolant temperature and different flux rate(s). This paper presents new creep correlations developed for ACR® fuel in 2009. It includes more recent world-wide data on the creep rates of Zircaloy-4. The additional data covers longer creep duration under various neutron flux rates and different temperatures and stress levels. The creep data and correlations are categorised as two types of materials: stress-relieved annealed (SRA) and recrystallised annealed (RXA). A time hardening and a strain hardening correlation are formulated respectively in this paper. The new correlations provide small statistical error, and agree well in overall predictions with the measured creep data for Zircaloy-4 sheaths. (author)

  2. CATHENA modelling of ACR-1000 fuel handling events

    The ACR-1000® Fuel Handling (FH) and Storage System provides on-power fuelling capability based on a proven CANDU® technology. It includes all aspects of fuel storage and handling, from the arrival of new fuel to the storage of irradiated fuel. The fuelling machine (FM) water system provides the cooling water to the fuelling machines during the normal operation. The FH emergency water system is a process circuit that is separate from the FM water system and the spent fuel transfer process system. The FH emergency water system operates at low pressures and is seismically and environmentally qualified. The paper presents a CATHENA model developed for the safety analysis of ACR-1000 FH events with the loss of cooling function in the fuelling machine water system. (author)

  3. ACR-1000: Product Update

    The ACR-1000 uses well-established, fundamental, CANDU design elements: core design with horizontal pressure tubes; simple efficient fuel bundle design; on-power refuelling and a separate low-pressure, low-temperature heavy-water moderator providing an inherent emergency heat sink. It includes adaptations for light-water coolant and low-enriched uranium fuel, and offers a compact core configuration and higher steam pressure for greater thermodynamic efficiency. The ACR-1000 links design with licensing, emphasizing operability and maintainability from the viewpoint of the customer-the utility operator. (authors)

  4. ACR-1000 - optimized plant for utility requirements

    The Generation III+ Advanced CANDU Reactor (ACR) is available in two sizes, the ACR-700 (750 MWe class) and ACR-1000 (1200 MWe class). Market forces (Canada, China, UK) have been pushing towards the larger ACR design, and AECL is now focusing its attention on the ACR-1000. The basic engineering program initiated for the ACR-700 provides the starting basis for the ACR-1000. We have established key ACR-1000 design objectives, input customer requirements and reviewed lessons learned from ACR-700 development and market feedback. The basic design of the ACR-1000 is virtually identical with that of the ACR-700, operating at similar reactor coolant system pressures and temperatures. This paper focuses on key ACR-1000 features: 1200 MWe output, robust design with passive resistance to severe accidents, reactor characteristics for operational safety and reliability, improved CANFLEX ACR fuel, a compact core with lattice pitch and reactor face design to enhance maintainability and inspection, and 'Smart' systems to permit on-line monitoring and diagnostics. (author)

  5. ACR-1000: Enhanced response to severe accidents

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

  6. ACR-1000: Operator - based development

    Atomic Energy of Canada Limited (AECL) has adapted the successful features of CANDU* reactors to establish Generation III+ Advanced CANDU ReactorTM (ACRTM) technology. The ACR-1000TM nuclear power plant is an evolutionary product, starting with the strong base of CANDU reactor technology, coupled with thoroughly-demonstrated innovative features to enhance economics, safety, operability and maintainability. The ACR-1000 benefits from AECL's continuous-improvement approach to design, that enabled the traditional CANDU 6 product to compile an exceptional track record of on-time, on budget product delivery, and also reliable, high capacity-factor operation. The ACR-1000 engineering program has completed the basic plant design and has entered detailed pre-project engineering and formal safety analysis to prepare the preliminary (non-project-specific) safety case. The engineering program is strongly operator-based, and encompasses much more than traditional pre-project design elements. A team of utility-experienced operations and maintenance experts is embedded in the engineering team, to ensure that all design decisions, at the system and the component level, are taken with the owner-operator interest in mind. The design program emphasizes formal review of operating feedback, along with extensive operator participation in program management and execution. Design attention is paid to layout and access of equipment, to component and material selection, and to ensuring maximum ability for on-line maintenance. This enables the ACR-1000 to offer a three-year interval between scheduled maintenance outages, with a standard 21-day outage duration. SMART CANDUTM technology allows on-line monitoring and diagnostics to further enhance plant operation. Modules of the Advanced CANDU SMART technologies are already being back-fitted to current CANDU plants. As well as reviewing the ACR-1000 design features and their supporting background, the paper describes the status of main program

  7. Methodology for fission product release calculations during an ACR-1000 end-fitting failure event

    The ACR-1000® reactor enhances and retains the proven features of the CANDU® design such as the concept of the horizontal fuel channel core. At each end of a fuel channel, there is an end-fitting incorporating a feeder connection through which pressurized coolant enters and leaves the fuel channel, where 12 fuel bundles are inserted. The safety analysis cases include postulated end-fitting failure events to assess the fission product releases from all fuel bundles which would be ejected out of the channel and oxidized in the air-steam environment under decay power. This paper presents the methodology used in assessing the fuel behaviour and the fission product releases during a postulated end-fitting failure in an ACR-1000 reactor. After the end-fitting failure, the 12 fuel bundles are ejected out of the channel and drop onto the fuelling machine vault floor. The fuel bundles are likely heavily damaged by impact and would break into small clusters of elements or fragments. To calculate the fission product releases from an individual fragment, the transient fuel temperature is numerically solved by differential heat equations; the air oxidation model is chosen for the event accordingly; and the fission product inventory and releases are estimated by computer codes ORIGEN-S, CATHENA, ELESTRES and SOURCE-IST. Finally, the total fission product releases from all fragments into containment are calculated. This methodology has been developed for ACR-1000 safety analysis, which is also applicable to CANDU. With the new methodology, the transient releases from up to 150 fission products can be estimated as detail as in fragment. In this paper, a sample calculation is also provided to show the application of the methodology in ACR-1000 safety analysis for end-fitting failure. (author)

  8. ACR-1000 - Designed for constructability

    Full text: One of the key aspects to be considered in the delivery of a Nuclear Power Plant is the security of the construction schedule and the need for lower construction costs. Many industries are using skids, modules and prefabrications to enhance construction productivity, reduce schedules and thus reduce costs. The leaders in this regard have traditionally been in the off-shore oil and gas, chemical, refinery and ship building industries. The concept of using modules has been utilized in Nuclear Power Plant design and construction. Atomic Energy of Canada Limited (AECL) has had considerable success at the Qinshan Nuclear Power project in China with the use of modularization, which proved extremely effective in the ability to organize parallel construction activities and shortening the schedule. Extensive use has been made of skids and modules in Japan and this also has proven effective in shortening schedules in the construction of nuclear power plants. Secondary benefits of modularization and prefabrication include decreased site congestion and logistical issues, increased worker safety and better quality control of fabrication. Modules and prefabrication allow work to be shifted to areas where skilled trades are more readily available from a site where skilled trades are very limited. One of the objectives of the ACR-1000 project is to produce a design that allows for a very secure construction schedule. The construction method and strategy, consisting of extensive use of prefabrication and modularization was defined very early in the ACR-1000 conceptual phase of the layout and design process. This has been achieved through a constructability programme that integrates the civil design with site erection and module installation. This approach takes the concept of modularization to an entirely new level, in which the use of modules is built into the design from the start, rather than backfitting modular construction into a conventionally designed plant. This

  9. ACR-1000 Project - Licensing Opportunities and Challenges

    challenges that have been successfully overcome by both CNSC and AECL. By providing valuable feedback, AECL has worked actively in review of the existing applicable reactor regulations in Canada, and in development of new regulatory requirements and guides. In this regard, AECL provided constructive comments on several draft regulatory guides, and performed compliance self- assessment against all applicable regulations to be used with the new build in Canada including the IAEA safety requirements. As there is no legal process in Canada for design certification of nuclear power plant (NPP), AECL initiated a pre-project regulatory review of the ACR-1000 reactor design by the CNSC to confirm compliance with regulatory requirements and also incorporate regulatory feedback in the design process to minimize project risks in obtaining future construction and operating licences for NPPs in Canada. This pre-project review consists of two phases starting from April 1, 2008 and ending on August 30, 2009. Phase 1 ended in December 2008 and has concluded that at an overall level the ACR-1000 design intent is compliant with the CNSC regulatory requirements and meets the expectations for new nuclear power plants in Canada. This conclusion is expected to be further confirmed during the Phase 2 review that currently is ongoing. Phase 2 will go into further detail with a focus on identifying whether or not any potential fundamental barriers to licensing the design in Canada. This phase involves review of 16 topical areas: 1. Defence in Depth, Classification of Structures, Systems and Components and Regulatory Dose Limits; 2. Reactor Physics Aspects of Nuclear Design; 3. Fuel Mechanical and Thermalhydraulics Design; 4. Reactor Control System; 5. Shutdown Means; 6. Emergency and Long Term Core Cooling, Emergency Feedwater System; 7. Containment and Reactor Auxiliary Building; 8. Safety Analysis; 9. Heat Transport System Pressure Boundary; 10. Fire Protection; 11. Radiation Protection; 12. Out

  10. A CANDU-6 versus ACR-1000 SDS1 performance comparison during some LOCA scenarios

    According to the Romanian Nuclear Strategy, the third and fourth units of the Cernavoda NPP will be commissioned by 2015. Improvements in operation and safety are expected to be applied for these CANDU-6 based units. On the other side, the need for innovation determined AECL to promote the ACR -1000 - an evolutionary Generation III+ power reactor design and a necessary step towards Generation IV inherently safe nuclear energy systems. CANDU-6 is recognized for having two independent fully capable shutdown systems. ACR-1000 also benefits for this strong safety feature. Two major achievements i.e. using of light water as coolant and using Low Enriched Uranium (LEU) as fuel in a compact heavy water moderated lattice allowed the obtaining of a slightly negative Coolant Void Reactivity (CVR) for the first time in a CANDU-type reactor. The main goal of the paper is to compare the response of SDS1 action during some LOCAs supposed to take place both in CANDU-6 and ACR-1000 reactors. In the considered scenarios, the initiation event was a Rupture of the Inlet Header (RIH) of 15, 25 or 35%. The analyses were performed using the point-kinetics approximation method implemented in the DIREN code - a 3D diffusion tool developed in INR Pitesti. The CANDU-6 core model is based on as-built data from Cernavoda Unit 1, while the ACR-1000 DIREN core model was recently developed during the PhD stage of the main author. The SOR reactivities, flux amplitude, maximum channel and bundle powers were the key parameters pursued in analyses. The results emphasized the net ACR-1000 safety improvement gained from its design innovations. (authors)

  11. Fuel bundle

    This patent describes a method of forming a fuel bundle of a nuclear reactor. The method consists of positioning the fuel rods in the bottom plate, positioning the tie rod in the bottom plate with the key passed through the receptacle to the underside of the bottom plate and, after the tie rod is so positioned, turning the tie rod so that the key is in engagement with the underside of the bottom plate. Thereafter mounting the top plate is mounted in engagement with the fuel rods with the upper end of the tie rod extending through the opening in the top plate and extending above the top plate, and the tie rod is secured to the upper side of sid top plate thus simultaneously securing the key to the underside of the bottom plate

  12. ACR1000 - Enhanced Safety for the Marketplace

    Building CANDU 6 units at Qinshan Phase III in China is contributed to the impressive schedule accomplishments. The ACR1000 is a 1200 MWe-class Generation III+ nuclear power plant with a 60-year design life. It is a light-water-cooled, heavy-water-moderated pressure-tube reactor, which has evolved from the well-established Candu line. It retains basic, proven, CANDU design features while incorporating innovations and state-of-the-art technologies to optimize safety, operation, performance and economics. These technical improvements, along with system simplifications and advancements in project engineering, manufacturing, and construction, result in a reduced capital cost and construction schedule, while enhancing the inherent safety and operating performance of the ACR1000 design. Maximum use of modularization and 'open-top', parallel construction - which have already been demonstrated at the Qinshan Phase III units, both delivered under budget and ahead of schedule - are key to AECL's ACR1000 project model. The ACR1000 has been chosen for generic design assessment in the United Kingdom. Additionally, there are active ACR1000 initiatives in Canada - in Ontario, where the ACR1000 has been short-listed for new build-Alberta and New Brunswick

  13. ACR-1000TM - advanced Candu reactor design

    Atomic Energy of Canada Limited (AECL) has developed the Advanced CANDU ReactorTM- 1000 (ACR-1000TM) as an evolutionary advancement of the current CANDU 6TM reactor. This evolutionary advancement is based on AECL's in-depth knowledge of CANDU structures, systems, components and materials, gained during 50 years of continuous construction, engineering and commissioning, as well as on the experience and feedback received from operators of CANDU plants. The ACR design retains the proven strengths and features of CANDU reactors, while incorporating innovations and state-of-the-art technology. These innovations improve economics, inherent safety characteristics, and performance, while retaining the proven benefits of the CANDU family of nuclear power plants. The Canadian nuclear reactor design evolution that has reached today's stage represented by the ACR-1000, has a long history dating back to the early 1950's. In this regard, Canada is in a unique situation, shared only by a very few other countries, where original nuclear power technology has been invented and further developed. The ACR design has been reviewed by domestic and international regulatory bodies, and has been given a positive regulatory opinion about its licensability. The Canadian regulator, the Canadian Nuclear Safety Commission (CNSC) completed the Phase 1 and Phase 2 pre-project design reviews in December 2008 and August 2009, respectively, and concluded that there are no fundamental barriers to licensing the ACR-1000 design in Canada. The final stage of the ACR-1000 design is currently underway and will be completed by fall of 2011, along with the final elements of the safety analyses and probabilistic safety analyses supporting the finalized design. The generic Preliminary Safety Analysis Report (PSAR) for the ACR-1000 was completed in September 2009. The PSAR demonstrates ACR-1000 safety case and compliance with Canadian and international regulatory requirements and expectations. (authors)

  14. Fuel bundle geometry and composition influence on coolant void reactivity reduction in ACR and CANDU reactors

    It is very well known that the CANDU reactor has positive Coolant Void Reactivity (CVR), which is most important criticisms about CANDU. The most recent innovations based on using a thin absorbent Hafnium shell in the central bundle element were successfully been applied to the Advanced CANDU Reactor (ACR) project. The paper's objective is to analyze elementary lattice cell effects in applying such methods to reduce the CVR. Three basic fuel designs in their corresponding geometries were chosen to be compared: the ACR-1000TM, the RU-43 (developed in INR Pitesti) and the standard CANDU fuel. The bundle geometry influence on void effect was also evaluated. The WIMS calculations proved the Hafnium absorber suitability (in the latest 'shell design') to achieve the negative CVR target with great accuracy for the ACR-1000 fuel bundle design than for the other two projects. (authors)

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

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

  16. candu fuel bundle fabrication

    This paper describes works on CANDU fuel bundle fabrication in the Fuel Fabrication Development and Testing Section (FFDT) of AECL's Chalk River Laboratories. This work does not cover fuel design, pellet manufacturing, Zircaloy material manufacturing, but cover the joining of appendages to sheath tube, endcap preparation and welding, UO2 loading, end plate preparation and welding, and all inspections required in these steps. Materials used in the fabrication of CANDU fuel bundle are: 1)Ceramic UO2 Pellet 2)Zircaloy -4. Fuel Bundle Structural Material 3) Others (Zinc stearate, Colloidal graphite, Beryllium and Heium). Th fabrication of fuel element consist of three process: 1)pellet loading into the sheats, 2) endcap welding, and 3) the element profiling. Endcap welds is tested by metallography and He leak test. The endcaps of the elements are welded to the end plates to form the 37- element bundle assembly

  17. ACR-1000 pre-project regulatory review progress

    The ACR-1000 design developed by Atomic Energy of Canada Limited (AECL) is a 1200 MWe-class light-water-cooled, heavy-water-moderated pressure-tube reactor, which has evolved from the well-established CANDU line of reactors. The ACR-1000 design retains the basic, proven, CANDU design features while incorporating innovations and state-of-the-art technologies to ensure fully competitive safety, operation, performance and economics. Improvements include greater operating and safety margins plus adherence and compliance with the latest safety objectives of designing with due consideration to external events and risk assessment. AECL initiated a pre-project regulatory review of the ACR-1000 reactor design by the Canadian Nuclear Safety Commission (CNSC) to confirm compliance with regulatory requirements and also incorporate regulatory feedback in the design process to minimize regulatory risks in obtaining construction and operating licences. Regulatory pre-project reviews have also been conducted earlier in the UK and US to ensure that the ACR design is compliant with international regulatory requirements. (author)

  18. Advanced Fuel Bundles for PHWRS

    The fuel used by NPCIL presently is natural uranium dioxide in the form of 19- element fuel bundles for 220 MWe PHWRs and 37-element fuel bundles for the TAPP-3&4 540 MWe units. The new 700 MWe PHWRs also use 37-element fuel bundles. These bundles are of short 0.5 m length of circular geometry. The cladding is of collapsible type made of Zircaloy-4 material. PHWRs containing a string of short length fuel bundles and the on-power refueling permit flexibility in using different advanced fuel designs and in core fuel management schemes. Using this flexibility, alternative fuel concepts are tried in Indian PHWRs. The advances in PHWR fuel designs are governed by the desire to use resources other than uranium, improve fuel economics by increasing fuel burnup and reduce overall spent nuclear fuel waste and improve reactor safety. The rising uranium prices are leading to a relook into the Thorium based fuel designs and reprocessed Uranium based and Plutonium based MOX designs and are expected to play a major role in future. The requirement of synergism between different type of reactors also plays a role. Increase in fuel burnup beyond 15 000 MW∙d/TeU in PHWRs, using higher fissile content materials like slightly enriched uranium, Mixed Oxide and Thorium Oxide in place of natural uranium in fuel elements, was studied many PHWR operating countries. The work includes reactor physics studies and test irradiation in research reactors and power reactors. Due to higher fissile content these bundles will be capable of delivering higher burnup than the natural uranium bundles. In India the fuel cycle flexibility of PHWRs is demonstrated by converting this type of technical flexibility to the real economy by irradiating these different types of advanced fuel materials namely Thorium, MOX, SEU, etc. The paper gives a review of the different advanced fuel design concepts studied for Indian PHWRs. (author)

  19. Locking means for fuels bundles

    A nuclear power reactor fuel bundle is described which has a plurality of fuel rods disposed between two end plates positioned by tie rods extending therebetween. The assembled bundle is secured by one or more locking forks which pass through slots in the tie rod ends. Springs mounted on the fuel rods and tie rods are compressed by assembling the bundle and forcing one end plate against the locking fork to maintain the fuel rods and tie rods in position between the end plates. Downward pressure on the end plate permits removal of the locking fork so that the end plates may be removed, thus giving access to the fuel rods. This construction facilitates disassembly of an irradiated fuel bundle under water

  20. CANFLEX fuel bundle impact test

    This document outlines the test results for the impact test of the CANFLEX fuel bundle. Impact test is performed to determine and verify the amount of general bundle shape distortion and defect of the pressure tube that may occur during refuelling. The test specification requires that the fuel bundles and the pressure tube retain their integrities after the impact test under the conservative conditions (10 stationary bundles with 31kg/s flow rate) considering the pressure tube creep. The refuelling simulator operating with pneumatic force and simulated shield plug were fabricated and the velocity/displacement transducer and the high speed camera were also used in this test. The characteristics of the moving bundle (velocity, displacement, impacting force) were measured and analyzed with the impact sensor and the high speed camera system. The important test procedures and measurement results were discussed as follows. 1) Test bundle measurements and the pressure tube inspections 2) Simulated shield plug, outlet flange installation and bundle loading 3) refuelling simulator, inlet flange installation and sensors, high speed camera installation 4) Perform the impact test with operating the refuelling simulator and measure the dynamic characteristics 5) Inspections of the fuel bundles and the pressure tube. (author). 8 refs., 23 tabs., 13 figs

  1. CANFLEX fuel bundle impact test

    Chang, Seok Kyu; Chung, C. H.; Park, J. S.; Hong, S. D.; Kim, B. D.

    1997-08-01

    This document outlines the test results for the impact test of the CANFLEX fuel bundle. Impact test is performed to determine and verify the amount of general bundle shape distortion and defect of the pressure tube that may occur during refuelling. The test specification requires that the fuel bundles and the pressure tube retain their integrities after the impact test under the conservative conditions (10 stationary bundles with 31kg/s flow rate) considering the pressure tube creep. The refuelling simulator operating with pneumatic force and simulated shield plug were fabricated and the velocity/displacement transducer and the high speed camera were also used in this test. The characteristics of the moving bundle (velocity, displacement, impacting force) were measured and analyzed with the impact sensor and the high speed camera system. The important test procedures and measurement results were discussed as follows. 1) Test bundle measurements and the pressure tube inspections 2) Simulated shield plug, outlet flange installation and bundle loading 3) refuelling simulator, inlet flange installation and sensors, high speed camera installation 4) Perform the impact test with operating the refuelling simulator and measure the dynamic characteristics 5) Inspections of the fuel bundles and the pressure tube. (author). 8 refs., 23 tabs., 13 figs.

  2. Defueled channel experiments in ZED-2 in support of ACR-1000 ROP analysis

    Defueled channel experiments were performed in ZED-2 to help resolve discrepancies between calculated flux detector response during refueling in ACR-1000 according the reactor codes RFSP and MCNP. The data produced from these experiments was later used in a separate Regional-Over-Power (ROP) analysis to verify MCNP and RFSP neutron response predictions during refueling. These experiments provided information on thermal flux distributions interior and exterior to a fueled and defueled channel; and on epithermal absolute flux distributions exterior to the same channel. Critical height and moderator temperature data for fueled and defueled channel conditions were also measured. In addition, standard platinum-clad Inconel Self-Powered Detector (SPD) performance data was obtained. The following reactor physics and SPD parameters were measured in these experiments: C Radial flux distribution inside the channel of interest (fueled and defueled), C Radial flux distribution outside the channel of interest (fueled and defueled), C Epithermal radial flux distribution outside the channel of interest (fueled and defueled), and C SPD response parallel to and normal to the channel of interest (fueled and defueled).

  3. Computational fluid dynamics model for liquid poison injection in the ACR-1000 design

    The Advanced CANDU Reactor (ACR-1000) Shutdown System 2 is capable of quickly rendering the reactor core subcritical by injecting a neutron absorbing solution (poison) into the heavy water moderator via injection nozzles. A Computational Fluid Dynamics (CFD) model has been developed to simulate the poison injection into the moderator. This paper presents the model development and preliminary results to demonstrate its feasibility to the ACR-1000 design. The CFD model has been validated against the test data from the CANDU 6 LISS test. Validation tests based on the ACR-1000 design are underway, in which the poison concentration distribution will be measured. (author)

  4. Best Available Technique (BAT) assessment applied to ACR-1000 waste and heavy water management systems

    The ACR-1000 design is the next evolution of the proven CANDU reactor design. One of the key objectives for this project was to systematically apply the As Low As Reasonably Achievable (ALARA) principle to the reactor design. The ACR design team selected the Best Available Technique (BAT) assessment for this purpose to document decisions made during the design of each ACR-1000 waste and heavy water management systems. This paper describes the steps in the BAT assessment that has been applied to the ACR-1000 design. (author)

  5. Static stress analysis of CANFLEX fuel bundles

    The static stress analysis of CANFLEX bundles is performed to evaluate the fuel structural integrity during the refuelling service. The structure analysis is carried out by predicting the drag force, stress and displacements of the fuel bundle. By the comparison of strength tests and analysis results, the displacement values are well agreed within 15%. The analysis shows that the CANFLEX fuel bundle keep its structural integrity. 24 figs., 6 tabs., 12 refs. (Author) .new

  6. Hydraulic characteristics of HANARO fuel bundles

    Cho, S.; Chung, H. J.; Chun, S. Y.; Yang, S. K.; Chung, M. K. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents the hydraulic characteristics measured by using LDV (Laser Doppler Velocimetry) in subchannels of HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids, which are arranged in cylindrical configuration. The effects of the spacer grids on the turbulent flow were investigated by the experimental results. Pressure drops for each component of the fuel bundle were measured, and the friction factors of fuel bundle and loss coefficients for the spacer grids were estimated from the measured pressure drops. Implications regarding the turbulent thermal mixing were discussed. Vibration test results measured by using laser vibrometer were presented. 9 refs., 12 figs. (Author)

  7. CANDU fuel bundle skin friction factor

    Single-phase, incompressible fluid flow skin friction factor correlations, primarily for CANDU 37-rod fuel bundles, were reviewed. The correlations originated from curve-fits to flow test data, mostly with new fuel bundles in new pressure tubes (flow tubes), without internal heating. Skin friction in tubes containing fuel bundles (noncircular flow geometry) was compared to that in equivalent diameter smooth circular tubes. At Reynolds numbers typical of normal flows in CANDU fuel channels, the skin friction in tubes containing bundles is 8 to 15% higher than in equivalent diameter smooth circular tubes. Since the correlations are based on scattered results from measurements, the skin friction with bundles may be even higher than indicated above. The information permits over- or under-prediction of the skin friction, or choosing an intermediate value of friction, with allowance for surface roughnesses, in thermal-hydraulic analyses of CANDU heat transport systems. (author) 9 refs., 2 figs

  8. ACR-1000: Advanced I and C and IT systems to enhance operations and maintenance

    The Advanced CANDU Reactor ACR-1000 is a 1200-MWe-class Generation III+ nuclear power plant designed by Atomic Energy of Canada Limited (AECL). Its design is evolutionary, starting with a strong base of proven CANDU reactor technology coupled with thoroughly demonstrated innovative features to enhance economics, safety, operability and maintainability. Two key design strategies were to expand the Instrumentation and Control (I and C) and Information Technology (IT) systems, and improve Operations and Maintenance (O and M) capability. AECL has developed ACR-1000 I and C and IT systems, including SMART CANDU, to improve the timeliness, the quality and the integration of the information made available to plant operators and engineers. These systems use automatic data mining and present organized and analyzed data to operators and engineers to facilitate diagnostics and reduce mental burden - thus increasing the ability to make proactive and informed decisions affecting plant operation. Major advances have also been made in designing and ACR-1000 control room itself. Additionally, direct feedback from CANDU plant operators on enhancing the ACR-1000's operability has permitted optimization of on-line maintenance and facilitated and reduced off-line maintenance. New features have been designed into the plant to reduce operating risk and reduce costs. This paper outlines how IT and O and M advances have enabled the ACR-1000 to meet and exceed performance and operability targets. (author)

  9. Bringing the CANFLEX fuel bundle to market

    CANFLEX is a 43-element CANDU fuel bundle, under joint development by AECL and KAERI, to facilitate the use of various advanced fuel cycles in CANDU reactors through the provision of enhanced operating margins. The bundle uses two element diameters (13.5 and 11.5 mm ) to reduce element ratings by 20%, and includes the use of critical-heat-flux (CHF) enhancing appendages to increase the minimum CHF ratio or dryout margin of the bundle. Test programs are underway to demonstrate: the irradiation behaviour, hydraulic characteristics and reactor physics properties of the bundle, along with a test program to demonstrate the ability of the bundle to be handled by CANDU-6 fuelling machines. A fuel design manual and safety analysis reports have been drafted, and both analyses, plus discussions with utilities are underway for a demonstration irradiation in a CANDU-6 reactor. (author)

  10. Nuclear fuel bundle disassembly and assembly tool

    A nuclear power reactor fuel bundle is described which has a plurality of tubular fuel rods disposed in parallel array between two transverse tie plates. It is secured against disassembly by one or more locking forks which engage slots in tie rods which position the transverse plates. Springs mounted on the fuel and tie rods are compressed when the bundle is assembled thereby maintaining a continual pressure against the locking forks. Force applied in opposition to the springs permits withdrawal of the locking forks so that one tie plate may be removed, giving access to the fuel rods. An assembly and disassembly tool facilitates removal of the locking forks when the bundle is to be disassembled and the placing of the forks during assembly of the bundle. (U.S.)

  11. Assembly mechanism for nuclear fuel bundles

    A description is given of a nuclear power reactor fuel bundle having tie rods fastened to a lower tie plate and passing through openings in the upper tie plate with the assembled bundle secured by rotatable locking sleeves which engage slots provided in the upper tie plate. Pressure exerted by helical springs mounted around each of the fuel rods urge the upper tie plate against the locking sleeves. The bundle may be disassembled after depressing the upper tie plate and rotating the locking sleeves to the unlocked position

  12. In-pool damaged fuel bundle recovery

    While preparing to rerack the Oyster Creek Nuclear Generating Station, GPU Nuclear had need to move a damaged fuel bundle. This bundle had no upper tie plate and could not be moved in the normal manner. GPU Nuclear formed a small, dedicated project team to disassemble, package, and move this damaged bundle. The team was composed of key personnel from GPU Nuclear Fuels Projects, OCNGS Operations and Proto-Power/Bisco, a specialty contractor who has fuel bundle reconstitution and rod consolidation experience, remote tooling, underwater video systems and experienced technicians. Proven tooling, clear procedures and a simple approach were important, but the key element was the spirit of teamwork and leadership exhibited by the people involved. In spite of several emergent problems which a task of this nature presents, this small, close knit utility/vendor team completed the work on schedule and within the exposure and cost budgets

  13. In-pool damaged fuel bundle recovery

    While preparing to rerack the Oyster Creek Nuclear Generating Station, GPU Nuclear had need to move a damaged fuel bundle. This bundle had no upper tie plate and could not be moved in the normal manner. GPU Nuclear formed a small, dedicated project team to disassemble, package and move this damaged bundle. The team was composed of key personnel from GPU Nuclear Fuels Projects, OCNGS Operations and Proto-Power / Bisco, a specialty contractor who has fuel bundle reconstitution and rod consolidation experience, remote tooling, underwater video systems and experienced technicians. Proven tooling, clear procedures and a simple approach were important, but the key element was the spirit of teamwork and leadership exhibited by the people involved

  14. CANFLEX fuel bundle strength tests (test report)

    Chang, Seok Kyu; Chung, C. H.; Kim, B. D.

    1997-08-01

    This document outlines the test results for the strength tests of the CANFLEX fuel bundle. Strength tests are performed to determine and verify the amount of the bundle shape distortion which is against the side-stops when the bundles are refuelling. There are two cases of strength test; one is the double side-stop test which simulates the normal bundle refuelling and the other is the single side-stop test which simulates the abnormal refuelling. the strength test specification requires that the fuel bundle against the side-stop(s) simulators for this test were fabricated and the flow rates were controlled to provide the required conservative hydraulic forces. The test rig conditions of 120 deg C, 11.2 MPa were retained for 15 minutes after the flow rate was controlled during the test in two cases, respectively. The bundle loading angles of number 13- number 15 among the 15 bundles were 67.5 deg CCW and others were loaded randomly. After the tests, the bundle shapes against the side-stops were measured and inspected carefully. The important test procedures and measurements were discussed as follows. (author). 5 refs., 22 tabs., 5 figs.

  15. CANFLEX fuel bundle strength tests (test report)

    This document outlines the test results for the strength tests of the CANFLEX fuel bundle. Strength tests are performed to determine and verify the amount of the bundle shape distortion which is against the side-stops when the bundles are refuelling. There are two cases of strength test; one is the double side-stop test which simulates the normal bundle refuelling and the other is the single side-stop test which simulates the abnormal refuelling. the strength test specification requires that the fuel bundle against the side-stop(s) simulators for this test were fabricated and the flow rates were controlled to provide the required conservative hydraulic forces. The test rig conditions of 120 deg C, 11.2 MPa were retained for 15 minutes after the flow rate was controlled during the test in two cases, respectively. The bundle loading angles of number 13- number 15 among the 15 bundles were 67.5 deg CCW and others were loaded randomly. After the tests, the bundle shapes against the side-stops were measured and inspected carefully. The important test procedures and measurements were discussed as follows. (author). 5 refs., 22 tabs., 5 figs

  16. Validation of MCNP and WIMS-AECL/DRAGON/RFSP for ACR-1000 applications

    Bromley, Blair P.; Adams, Fred P.; Zeller, Michael B.; Watts, David G.; Shukhman, Boris V.; Pencer, Jeremy [AECL - Chalk River Laboratories, Chalk River (Canada)

    2008-07-01

    This paper gives a summary of the validation of the reactor physics codes WIMS-AECL, DRAGON, RFSP and MCNP5, which are being used in the design, operation, and safety analysis of the ACR-1000{sup R}. The standards and guidelines being followed for code validation of the suite are established in CSA Standard N286.7-99 and ANS Standard ANS-19.3-2005. These codes are being validated for the calculation of key output parameters associated with various reactor physics phenomena of importance during normal operations and postulated accident conditions in an ACR-1000 reactor. Experimental data from a variety of sources are being used for validation. The bulk of the validation data is from critical experiments in the ZED-2 research reactor with ACR-type lattices. To supplement and complement ZED-2 data, qualified and applicable data are being taken from other power and research reactors, such as existing CANDU{sup R} units, FUGEN, NRU and SPERT research reactors, and the DCA critical facility. MCNP simulations of the ACR-1000 are also being used for validating WIMS-AECL/ DRAGON/RFSP, which involves extending the validation results for MCNP through the assistance of TSUNAMI analyses. Code validation against commissioning data in the first-build ACR-1000 will be confirmatory. The code validation is establishing the biases and uncertainties in the calculations of the WIMS-AECL/DRAGON/RFSP suite for the evaluation of various key parameters of importance in the reactor physics analysis of the ACR-1000. (authors)

  17. In-pile test of Qinshan PWR fuel bundle

    In-pile test of Qinshan Nuclear Power Plant PWR fuel bundle has been conducted in HWRR HTHP Test loop at CIAE. The test fuel bundle was irradiated to an average burnup of 25000 Mwd/tU. The authors describe the structure of (3 x 3-2) test fuel bundle, structure of irradiation rig, fuel fabrication, irradiation conditions, power and fuel burnup. Some comments on the in-pile performance for fuel bundle, fuel rod and irradiation rig were made

  18. TRIGA spent fuel bundles safe storage

    Negut, G.; Covaci, St. [Institute for Nuclear Research, Research Reactor Dept., Pitesti (Romania); Prisecaru, I.; Dupleac, D. [Bucharest Univ. Politehnica, Power and Nuclear Engineering Dept., Bucharest (Romania)

    2007-07-01

    TRIGA-SSR is a steady state research and material test reactor that has been in operation since 1980. The original TRIGA fuel was HEU (highly enriched uranium) with a U{sup 235} enrichment of 93 per cent. Almost all TRIGA HEU fuel bundles are now burned-up. Part of the spent fuel was loaded and transferred to US, in a Romania - DOE arrangement. The rest of the TRIGA fuel bundles have to be temporarily stored in the TRIGA facility. As the storage conditions had to be established with caution, neutron and thermal hydraulic evaluations of the storage conditions were required. Some criticality evaluations were made based on the SAR (Safety Analysis Report) data. Fuel constant axial temperature approximation effect is usual for criticality computations. TRIGA-SSR fuel bundle geometry and materials model for SCALE5-CSAS module allows the introduction of a fuel temperature dependency for the entire fuel active height, using different materials for each fuel bundle region. Previous RELAP5 thermal hydraulic computations for an axial and radial power distribution in the TRIGA fuel pin were done. Fuel constant temperature approximation overestimates pin factors for every core operating at high temperatures. From the thermal hydraulic point of view the worst condition of the storage grid occurs when the transfer channel is accidentally emptied of water from the pool, or the bundle is handled accidentally to remain in air. All the residual heat from the bundles has to be removed without fuel overheating and clad failure. RELAP5 computer code for residual heat removal was used in the assessment of residual heat removal. We made a couple of evaluations of TRIGA bundle clad temperatures in air cooling conditions, with different residual heat levels. The criticality computations have shown that the spent TRIGA fuel bundles storage grid is strongly sub-critical with k(eff) = 0.5951. So, there is no danger for a criticality accident for this storage grid type. The assessment is done

  19. TRIGA spent fuel bundles safe storage

    TRIGA-SSR is a steady state research and material test reactor that has been in operation since 1980. The original TRIGA fuel was HEU (highly enriched uranium) with a U235 enrichment of 93 per cent. Almost all TRIGA HEU fuel bundles are now burned-up. Part of the spent fuel was loaded and transferred to US, in a Romania - DOE arrangement. The rest of the TRIGA fuel bundles have to be temporarily stored in the TRIGA facility. As the storage conditions had to be established with caution, neutron and thermal hydraulic evaluations of the storage conditions were required. Some criticality evaluations were made based on the SAR (Safety Analysis Report) data. Fuel constant axial temperature approximation effect is usual for criticality computations. TRIGA-SSR fuel bundle geometry and materials model for SCALE5-CSAS module allows the introduction of a fuel temperature dependency for the entire fuel active height, using different materials for each fuel bundle region. Previous RELAP5 thermal hydraulic computations for an axial and radial power distribution in the TRIGA fuel pin were done. Fuel constant temperature approximation overestimates pin factors for every core operating at high temperatures. From the thermal hydraulic point of view the worst condition of the storage grid occurs when the transfer channel is accidentally emptied of water from the pool, or the bundle is handled accidentally to remain in air. All the residual heat from the bundles has to be removed without fuel overheating and clad failure. RELAP5 computer code for residual heat removal was used in the assessment of residual heat removal. We made a couple of evaluations of TRIGA bundle clad temperatures in air cooling conditions, with different residual heat levels. The criticality computations have shown that the spent TRIGA fuel bundles storage grid is strongly sub-critical with k(eff) = 0.5951. So, there is no danger for a criticality accident for this storage grid type. The assessment is done for

  20. Assembly mechanism for nuclear fuel bundles

    This invention relates to an assembly mechanism for nuclear power reactor fuel bundles using a novel, simple and inexpensive means. The mechanism is readily operable remotely, avoids separable parts and is applicable to fuel assemblies in which the upper tie plate is rigidly mounted on the tie rods which hold it in place. (UK)

  1. Development of nuclear fuel. Development of CANDU advanced fuel bundle

    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)

  2. CANFLEX - an advanced fuel bundle for CANDU

    The performance of CANDU pressurized heavy-water reactors, in terms of lifetime load factors, is excellent. More than 600 000 bundles containing natural-uranium fuel have been irradiated, with a low defect rate; reactor unavailability due to fuel incidents is typically zero. To maintain and improve CANDU's competitive position, Atomic Energy of Canada Limited (AECL) has an ongoing program comprising design, safety and availability improvements, advanced fuel concepts and schemes to reduce construction time. One key finding is that the introduction of slightly-enriched uranium (SEU, less than 1.5 wt% U-235 in U) offers immediate benefits for CANDU, in terms of fuelling and back-end disposal costs. The use of SEU places more demands on the fuel because of extended burnup, and an anticipated capability to load-follow also adds to the performance requirements. To ensure that the duty-cycle targets for SEU and load-following are achieved, AECL is developing a new fuel bundle, termed CANFLEX (CANdu FLEXible), where flexible refers to the versatility of the bundle with respect to operational and fuel-cycle options. Though the initial purpose of the new 43-element bundle is to introduce SEU into CANDU, CANFLEX is extremely versatile in its application, and is compatible with other fuel cycles of interest: natural uranium in existing CANDU reactors, recycled uranium and mixed-oxides from light-water reactors, and thoria-based fuels. Capability with a variety of fuel cycles is the key to future CANDU success in the international market. The improved performance of CANFLEX, particularly at high burnups, will ensure that the full economic benefits of advanced fuels cycles are achieved. A proof-tested CANFLEX bundle design will be available in 1993 for large-scale commercial-reactor demonstration

  3. SEU43 fuel bundle shielding analysis during spent fuel transport

    Margeanu, C. A.; Ilie, P.; Olteanu, G. [Inst. for Nuclear Research Pitesti, No. 1 Campului Street, Mioveni 115400, Arges County (Romania)

    2006-07-01

    The basic task accomplished by the shielding calculations in a nuclear safety analysis consist in radiation doses calculation, in order to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper investigates the effects induced by fuel bundle geometry modifications on the CANDU SEU spent fuel shielding analysis during transport. For this study, different CANDU-SEU43 fuel bundle projects, developed in INR Pitesti, have been considered. The spent fuel characteristics will be obtained by means of ORIGEN-S code. In order to estimate the corresponding radiation doses for different measuring points the Monte Carlo MORSE-SGC code will be used. Both codes are included in ORNL's SCALE 5 programs package. A comparison between the considered SEU43 fuel bundle projects will be also provided, with CANDU standard fuel bundle taken as reference. (authors)

  4. Telescope sipping - pinpointing leaking fuel bundles

    Given the top priority operators of nuclear power plants assign to safety, even the slightest sign of damage to the fuel assemblies has to be carefully monitored and analyzed. The detection of leaking fuel bundles also plays an important role in ensuring good availability and economy for the plants. ABB Atom has developed a new, highly accurate method, called 'telescope sipping', for identifying defective fuel assemblies. (orig.)

  5. Advances in the ACR-1000 reactor regulating system and reactor control

    Advances in the control of the ACR-1000 reactor are presented. The ACR-1000 Reactor Regulating System's (RRS) capability to maintain reactor power at its set point, counteract zonal power deviations, initiate setback as required, and effectively control operational maneuvers including power load-cycling is demonstrated. Three fast core transients and a long Load Cycling transient are presented. For simulations of the fast transients a dynamic RRS Simulation Package (RRS-SP) was developed, where the core neutron kinetics calculations (*CERBERUS module of RFSP) were coupled to a thermal hydraulic code (CATHENA) at every time step. A quasi-static approach was used to demonstrate the RRS performance in the Load Cycling transient that covers five consecutive daily cycles followed by a 2-day weekend cycle. (author)

  6. Regulatory assessment of effectiveness of ACR-1000 emergency core cooling system

    The paper presents the regulatory approach for assessment of the Advanced CANDU Reactor (ACR)-1000 Large Loss of Coolant Accident (LOCA) Emergency Core Cooling (ECC) effectiveness, describes the rationale for the selection of sensitivity cases and discusses the results of the simulations for 50% Pump Suction Break (PSB). The separate in-house simulations strengthened the CNSC staff knowledge about the ACR-1000 design and the modeling methodology. The review of representation of plant systems and plant behavior indicated no major issues. The selected accident scenarios and the limited scope sensitivity cases conducted by the CNSC staff, indicated that, overall, the ECC performance showed small sensitivity to the parameters and assumptions considered for investigation. (author)

  7. Strategy for 100-year life of the ACR-1000 concrete containment structure

    The purpose of this paper is to present the Plant Life Management (PLiM) strategy for the concrete containment structure of the ACR-1000 (Advanced CANDU Reactor) designed by AECL. The ACR-1000 is designed for 100-year plant life including 60-year operating life and additional 40-year decommissioning period of time. The approach adopted for the PLiM strategy of the concrete containment structure is a preventive one, key areas being: 1) design methodology, 2) material performance and 3) life cycle management and ageing management program. In the design phase, in addition to strength and serviceability, durability is a major requirement during the service life and decommissioning phase of the ACR structure. Parameters affecting durability design include: a) concrete performance, b) structural application, and c) environmental conditions. Due to the complex nature of the environmental effects acting on structures during the service life of project, it is considered that true improved performance during the service life can be achieved by improving the material characteristics. Many recent innovations in advanced concrete materials technology have made it possible to produce modern concrete such as high-performance concrete with exceptional performance characteristics. In this paper, the PLiM strategy for the ACR-1000 concrete containment is presented. In addition to addressing the design methodology and material performance areas, a systematic approach for ageing management program for the concrete containment structure is presented. (author)

  8. Fuel temperature characteristics of the 37-element and CANFLEX fuel bundle

    This report describes the fuel temperature characteristics of CANFLEX fuel bundles and 37-element fuel bundles for a different burnup of fuel. The program was consisted for seeking the fuel temperature of fuel bundles of CANFLEX fuel bundles and 37-element fuel bundles by using the method in NUCIRC. Fuel temperature has an increasing pattern with the burnup of fuel for CANFLEX fuel bundles and 37-element fuel bundles. For all the case of burnup, the fuel temperature of CANFLEX fuel bundles has a lower value than that of 37-element fuel bundles. Especially, for the high power channel, the CANFLEX fuel bundles show a lower fuel temperature as much as about 75 degree, and the core averaged fuel temperature has a lower fuel temperature of about 50 degree than that of 37-element fuel bundles. The lower fuel temperature of CANFLEX fuel bundles is expected to enhance the safety by reducing the fuel temperature coefficient. Finally, for each burnup of CANFLEX fuel bundles and 37-element fuel bundles, the equation was present for predicting the fuel temperature of a bundle in terms of a coolant temperature and bundle power

  9. Monitoring defective CANDU fuel bundles

    In 2005, it was proposed that a passive substance such as Nanocrystals could be used to monitor and locate defect fuel elements in-core. The experimental goal was to determine if Nanocrystals could be used for this application. Originally nanocrystals tagging was suggested for current operational CANDU-600 fuel. Other methods, including noble gas tagging, are also being investigated. Moreover, the scope of the project has been extended to include the identification of Dysprosium-doped fuel in the new ACR fuel design. The purpose of this paper is to discuss the experimental progress made at RMC on this project. (author)

  10. Using Advanced Fuel Bundles in CANDU Reactors

    Improving the exit fuel burnup in CANDU reactors was a long-time challenge for both bundle designers and performance analysts. Therefore, the 43-element design together with several fuel compositions was studied, in the aim of assessing new reliable, economic and proliferation-resistant solutions. Recovered Uranium (RU) fuel is intended to be used in CANDU reactors, given the important amount of slightly enriched Uranium (~0.96% w/o U235) that might be provided by the spent LWR fuel recovery plants. Though this fuel has a far too small U235 enrichment to be used in LWR's, it can be still used to fuel CANDU reactors. Plutonium based mixtures are also considered, with both natural and depleted Uranium, either for peacefully using the military grade dispositioned Plutonium or for better using Plutonium from LWR reprocessing plants. The proposed Thorium-LEU mixtures are intended to reduce the Uranium consumption per produced MW. The positive void reactivity is a major concern of any CANDU safety assessment, therefore reducing it was also a task for the present analysis. Using the 43-element bundle with a certain amount of burnable poison (e.g. Dysprosium) dissolved in the 8 innermost elements may lead to significantly reducing the void reactivity. The expected outcomes of these design improvements are: higher exit burnup, smooth/uniform radial bundle power distribution and reduced void reactivity. Since the improved fuel bundles are intended to be loaded in existing CANDU reactors, we found interesting to estimate the local reactivity effects of a mechanical control absorber (MCA) on the surrounding fuel cells. Cell parameters and neutron flux distributions, as well as macroscopic cross-sections were estimated using the transport code DRAGON and a 172-group updated nuclear data library. (author)

  11. Laser cutting for dismantling of PHWR fuel bundles

    Detailed investigation was carried out on laser cutting of zircaloy-2 PHWR fuel pin bundles. Initially, trials were done to standardize ten parameters for cutting of tie plates to which individual fuel pins are welded in a bundle. Using these parameters, the tie plates were cut into several pieces so that each fuel pin is individually separated out from the bundle. (author)

  12. Assembly mechanism for nuclear fuel bundles

    In a nuclear power reactor fuel bundle having tie rods fastened to a lower tie plate and passing through openings in the upper tie plate, the assembled bundle is secured by locking lugs fixed to rotatable locking sleeves which engage the upper tie plate. Pressure exerted by helical springs mounted around each of the tie rods urge retaining lugs fixed to a retaining sleeve associated with respective tie rods into a position with respect to the locking sleeve to prevent accidental disengagement of the upper plate from the locking lugs. The bundle may be disassembled by depressing the retaining sleeves and rotating the locking lugs to the disengaged position, and then removing the upper tie plate

  13. Assembly mechanism for nuclear fuel bundles

    The invention relates to a nuclear power reactor fuel bundle of the type wherein several rods are mounted in parallel array between two tie plates which secure the fuel rods in place and are maintained in assembled position by means of a number of tie rods secured to both of the end plates. Improved apparatus is provided for attaching the tie rods to the upper tie plate by the use of locking lugs fixed to rotatable sleeves which engage the upper tie plate. (auth)

  14. Enthalpy and void distributions in subchannels of PHWR fuel bundles

    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)

  15. Effect of bundle size on BWR fuel bundle critical power performance

    Effect of the bundle size on the BWR fuel bundle critical power performance was studied. For this purpose, critical power tests were conducted with both 6 x 6 (36 heater rods) and 12 x 12 (144 heater rods) size bundles in the GE ATLAS heat transfer test facility located in San Jose, California. All the bundle geometries such as rod diameter, rod pitch and rod space design are the same except size of flow channel. Two types of critical power tests were performed. One is the critical power test with uniform local peaking pattern for direct comparison of the small and large bundle critical power. Other is the critical power test for lattice positions in the bundle. In this test, power of a group of four rods (2 x 2 array) in a lattice region was peaked higher to probe the critical power of that lattice position in the bundle. In addition, the test data were compared to the COBRAG calculations. COBRAG is a detailed subchannel analysis code for BWR fuel bundle developed by GE Nuclear Energy. Based on these comparisons the subchannel model was refined to accurately predict the data obtained in this test program, thus validating the code capability of handling the effects of bundle size on bundle critical power for use in the study of the thermal hydraulic performance of the future advance BWR fuel bundle design. The author describes the experimental portion of the study program

  16. Assembly mechanism for nuclear fuel bundles

    A method of securing a fuel bundle to permit easy remote disassembly is described. Fuel rods are held loosely between end plates, each end of the rods fitting into holes in the end plates. At the upper end of each fuel rod there is a spring pressing against the end plate. Tie rods are used to hold the end plates together securely. The lower end of each tie rod is screwed into the lower end plate; the upper end of each tie rod is attached to the upper end plate by means of a locking assembly described in the patent. In order to remove the upper tie plate during the disassembly process, it is necessary only to depress the tie plate against the pressure of the springs surrounding the fuel rods and then to rotate each locking sleeve on the tie rods from its locked to its unlocked position. It is then possible to remove the tie plate without disassembling the locking assembly. (LL)

  17. Analysis of the Bundle Duct Interaction using the FBR fuel pin bundle deformation analysis code 'BAMBOO'

    PNC has been developing a computer code 'BAMBOO' to analyze the wire spaced FBR fuel pin bundle deformation under the BDI (Bundle Duct Interaction) condition by means of the three dimensional F.E.M. This code analyzes fuel pins' bowing and oval deformations which are dominant deformation behaviors of the fuel pin bundle under the BDI condition. In this study the 'BAMBOO' code is validated on the out-of-pile compression test of the FBR bundle (compression test) by comparing the results of the code analysis with the compression test results, and the highly irradiated (≥2.1x1027 n/m2, E > 0.1 MeV) bundle deformation behaviors are investigated from the viewpoint of the similarity to those in the compression test based on the analytical results of the code. (1) The calculated pin-to-duct minimum clearances as a function of the BDI levels in the compression test analysis agree with the experimental values evaluated from the CT image analysis of the bundle cross-section in the compression test within ±0.2 mm. And the calculated values of the fuel pins' oval deformations agree with the experimental values based on the pin diameter measurements done after the compression test within ±0.05 mm. (2) By comparing the irradiation induced bundle deformation with the bundle deformation in the compression test based on the code analysis, it is confirmed that the changes of the pin-to-duct minimum clearances with the BDI levels show equivalent trends between the both bundle deformations. And in this code analysis of the irradiation induced bundle deformation, contact loads between the fuel pins and the pacer wires are extremely small (below 10 kgf) even at about 3 dw of the BDI level compared to those in the compression test analysis. (J.P.N.)

  18. Fabrication of PWR fuel assembly and CANDU fuel bundle

    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

  19. Design and fabrication of a remote fuel bundle welding system

    A remote fuel bundle welding system in the hot-cell was designed and fabricated. To achieve this, a preliminary investigation of a hands-on fuel fabrication outside the hot-cell was conducted with a consideration of the constraints caused by welding in the hot-cell. Some basic experiments were also carried out to improve the end-plate welding process for fuel bundle manufacturing. The resistance welding system using the end-plate welding was also improved. It was found that resistance welding was more suitable for joining and end-plate to end caps in the hot-cell. The optimum conditions for end-plate welding for remote operation were also obtained. Preliminary performances to improve the resistance welding process were also examined, and the resistance welding process was determined to be the best in the hot-cell environment for fuel bundle manufacturing. The greatest advantage of fuel bundle welding system would be a qualified process for resistance welding in which there is extensive production experience. This paper presents an outline of the developed welding system for fuel bundle manufacturing and reviews the conceptual design of remote welding system using a master-slave manipulator. The design of a remote welding system using the 3-dimensional modeling method was also designed. Furthermore the mechanical considerations and the mock-up simulation test were described. Finally, its performance test results were presented for a mock-up of a remote fuel bundle welding system. (Author)

  20. Fuel rod bundles proposed for advanced pressure tube nuclear reactors

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

  1. Spring and stop assembly for nuclear fuel bundle

    A removable spring and stop assembly is described for use with a nuclear fuel bundle in a nuclear reactor core. The assembly includes a bolt threaded through a top section of a stop member by which the assembly (and a flow channel) is secured to the fuel bundle, the adjacent end threads of the bolt. The stop member is upset or deformed by which the bolt is captured in the assembly. (U.S.)

  2. Filler metals for containers holding irradiated fuel bundles

    One of the procedures being considered for the disposal of Canadian deuterium uranium (CANDU) irradiated fuel bundles is to place the bundles in containers, fill the containers with metal, and place them underground. This investigation deals with the selection of the filler metal with particular reference to the reaction rate with, and bonding of the filler metal to, the fuel sheathing (Zircaloy 4) and potential container materials. Lead, zinc, and aluminium alloys were examined as potential filler metals. (U.K.)

  3. Bundle duct interaction studies for fuel assemblies

    It is known that the wire-wrapped rods and duct in an LMFBR are undergoing a gradual structural distortion from the initially uniform geometry under the combined effects of thermal expansion and irradiation induced swelling and creep. These deformations have a significant effect on flow characteristics, thus causing changes in thermal behavior such as cladding temperature and temperature distribution within a bundle. The temperature distribution may further enhance or retard irradiation induced deformation of the bundle. This report summarizes the results of the continuing effort in investigating the bundle-duct interaction, focusing on the need for the large development plant

  4. Analysis of CHF experiment data for finned fuel bundle

    The HANARO uses finned-element fuel bundles. For thermal-hydraulic safety analysis, used is the MATRA-h code which is a modified version of KAERI's MATRA-α. The subchannel analysis model was determined by using the in-core irradiation test results and hydraulic experiment results for fuel bundle. The validity of the analysis model was investigated by comparing the MATRA-h predictions with the experimental results from several bundle CHF tests. The comparison showed that the code predictions for the CHF power were very close to or less than the experimental results. Thus, it was confirmed that the subchannel analysis using MATRA-h is to be applicable to the prediction of CHF phenomenon in HANARO fuel bundle

  5. Interactive hypermedia training manual for spent-fuel bundle counters

    Spent-fuel bundle counters, developed by the Canadian Safeguards Support Program for the International Atomic Energy Agency, provide a secure and independent means of counting the number of irradiated fuel bundles discharged into the fuel storage bays at CANDU nuclear power stations. Paper manuals have been traditionally used to familiarize IAEA inspectors with the operation, maintenance and extensive reporting capabilities of the bundle counters. To further assist inspectors, an interactive training manual has been developed on an Apple Macintosh computer using hypermedia software. The manual uses interactive animation and sound, in conjunction with the traditional text and graphics, to simulate the underlying operation and logic of the bundle counters. This paper presents the key features of the interactive manual and highlights the advantages of this new technology for training

  6. Study Of The PWR Fuel Bundle Characteristic With Borated Water

    Study of the PWR fuel bundle characteristic with 2,4, 2,6, 2,8, 3,0, 3,2 and 3,4 enrichment also with borated water 150 and 200 ppm has been done. The fuel bundle contained 264 fuel elements and water (no fuel elements) are arranged as 17 x 17 matrix and 30,294 cm. The fuel bundle characteristic can be seen from their group constants and the infinite multiplication factor whether more or less than one. The fuel bundle parameters can be found from cell calculation with WIMS PC version program. From the cell calculation shown that the infinite multiplication factor of the fuel bundle with 2,4% enrichment and 200 ppm borated water is 1, 01672, its shown that infinite multiplication factor will less than one with increasing borated water more than 200 ppm. From these result if we would like to design the reactor core with 2,4% minimum enrichment then the maximum borated water is 200 ppm

  7. Assessing the impact of the 37M fuel bundle design on fuel safety parameters

    To improve the critical heat flux and margin to fuel dryout in aging CANDU nuclear generating stations, the 37-element bundle design '37R' fuel) has been modified by reducing the central fuel element diameter, producing the modified '37M' fuel bundle. The codes FACTARSS, ELESTRES, ELOCA-IST, and SOURCE have been used to compare fuel temperature, fission gas release, and element integrity in 37R and 37M fuel bundles for Bruce Power nuclear reactors. The assessment demonstrated that, relative to 37R fuel bundles, using 37M fuel bundles does not significantly impact the existing safety margins associated with fuel temperature, fission gas release, and element integrity during design basis accidents. (author)

  8. Chop-leach fuel bundle residues densification by melting

    Nelson, R.G.; Griggs, B.

    1976-11-01

    Two melting processes were investigated for the densification of fuel bundle residues: Industoslag melting and graphite crucible melting. The Industoslag process, with prior decontamination and sorting, can produce ingots of Zircaloy, stainless steel and Inconel of a quality suitable for refabrication and reuse. The process can also melt oxidized mixtures of fuel bundle residues for direct storage. Eutectic mixtures of these materials can be melted in graphite at temperatures of 1300/sup 0/C. Hydrogen absorption experiments with the zirconium-rich alloys show the alloys to be potential tritium reservoirs. 13 figures.

  9. Performance of candu-6 fuel bundles manufactured in romania nuclear fuel plant

    The purpose of this article is to present the performance of nuclear fuel produced by Nuclear Fuel Plant (N.F.P.) - Pitesti during 1995 - 2012 and irradiated in units U1 and U2 from Nuclear Power Plant (N.P.P.) Cernavoda and also present the Nuclear Fuel Plant (N.F.P.) - Pitesti concern for providing technology to prevent the failure causes of fuel bundles in the reactor. This article presents Nuclear Fuel Plant (N.F.P.) - Pitesti experience on tracking performance of nuclear fuel in reactor and strategy investigation of fuel bundles notified as suspicious and / or defectives both as fuel element and fuel bundle, it analyzes the possible defects that can occur at fuel bundle or fuel element and can lead to their failure in the reactor. Implementation of modern technologies has enabled optimization of manufacturing processes and hence better quality stability of achieving components (end caps, chamfered sheath), better verification of end cap - sheath welding. These technologies were qualified by Nuclear Fuel Plant (N.F.P.) - Pitesti on automatic and Computer Numerical Control (C.N.C.) programming machines. A post-irradiation conclusive analysis which will take place later this year (2013) in Institute for Nuclear Research Pitesti (the action was initiated earlier this year by bringing a fuel bundle which has been reported defective by pool visual inspection) will provide additional information concerning potential damage causes of fuel bundles due to manufacturing processes. (authors)

  10. Interconnection of bundled solid oxide fuel cells

    Brown, Michael; Bessette, II, Norman F; Litka, Anthony F; Schmidt, Douglas S

    2014-01-14

    A system and method for electrically interconnecting a plurality of fuel cells to provide dense packing of the fuel cells. Each one of the plurality of fuel cells has a plurality of discrete electrical connection points along an outer surface. Electrical connections are made directly between the discrete electrical connection points of adjacent fuel cells so that the fuel cells can be packed more densely. Fuel cells have at least one outer electrode and at least one discrete interconnection to an inner electrode, wherein the outer electrode is one of a cathode and and anode and wherein the inner electrode is the other of the cathode and the anode. In tubular solid oxide fuel cells the discrete electrical connection points are spaced along the length of the fuel cell.

  11. Study on Unigraphics Drawing Modeling Method for 37-Element and CANFLEX Fuel Bundle

    The CANFLEX bundle contains 43 elements of two different diameters. It has two rings of small diameter elements on the outside, and eight elements (with diameter slightly larger than those in the standard 37-Element bundle) in the center. This larger number of small diameter elements on the outside of the CANFLEX bundle enhances thermo-hydraulic capability, resulting in a higher power capability and an improvement in operating safety margins. As a Result of advanced fuel design for CANFLEX fuel bundles, components consisting of fuel bundles are more complicated. Hence, the detailed modeling of components is inevitable in order to analyze the fuel performance by computational fluid dynamics. In this report, the basic design of the advanced fuel for CANDU reactors was carried out and the methodology for the modeling of fuel bundle were described. Firstly, the components consisting of fuel bundles were separately modeled and saved with different file names. The final feature of fuel bundle was accomplished by an assembling process of components. Since this report developed the modeling methodology based on the Unigraphics program, the basic explanations for the software were given first, and the complete modeling of 37-elements and CANFLEX fuel bundles were provided. The components of CANFLEX fuel bundles were also compared with that of 37-elements fuel bundles. Although, in this report, the modeling methodology is applied only to 37-elements and CANFLEX fuel bundles, this methodology may be applicable to the newly designed fuel bundles which are to be developed in the future

  12. Study on Unigraphics Drawing Modeling Method for 37-Element and CANFLEX Fuel Bundle

    Jeon, Yu Mi; Park, Joo Hwan

    2010-03-15

    The CANFLEX bundle contains 43 elements of two different diameters. It has two rings of small diameter elements on the outside, and eight elements (with diameter slightly larger than those in the standard 37-Element bundle) in the center. This larger number of small diameter elements on the outside of the CANFLEX bundle enhances thermo-hydraulic capability, resulting in a higher power capability and an improvement in operating safety margins. As a Result of advanced fuel design for CANFLEX fuel bundles, components consisting of fuel bundles are more complicated. Hence, the detailed modeling of components is inevitable in order to analyze the fuel performance by computational fluid dynamics. In this report, the basic design of the advanced fuel for CANDU reactors was carried out and the methodology for the modeling of fuel bundle were described. Firstly, the components consisting of fuel bundles were separately modeled and saved with different file names. The final feature of fuel bundle was accomplished by an assembling process of components. Since this report developed the modeling methodology based on the Unigraphics program, the basic explanations for the software were given first, and the complete modeling of 37-elements and CANFLEX fuel bundles were provided. The components of CANFLEX fuel bundles were also compared with that of 37-elements fuel bundles. Although, in this report, the modeling methodology is applied only to 37-elements and CANFLEX fuel bundles, this methodology may be applicable to the newly designed fuel bundles which are to be developed in the future

  13. Coupling Systems of Five CARA Fuel Bundles for Atucha I

    This paper describe the mechanical design of two options for the coupling systems of five CARA fuel bundles, to be used in the Atucha I nuclear power plant. These systems will be hydraulic tested in a low pressure loop to know their hydraulic loss of pressure

  14. Canflex: A fuel bundle to facilitate the use of enrichment and fuel cycles in CANDU reactors

    The neutron economy of the CANDU reactor results in it being an ideal host for a number of resource-conserving fuel cycles, as well as a number of potential ''symbiotic'' fuel cycles, in which fuel discharged from light-water cooled reactors is recycled to extract the maximum energy from the residual fissile material before it is sent for disposal. The resource conserving fuel cycles include the natural-uranium, slightly-enriched-uranium and thorium fuel cycles. The ''LWR-symbiotic'' cycles include recovered uranium and various options for the direct use of spent LWR fuel in CANDU reactors. However, to achieve the maximum economic potential of these fuel-cycle options requires irradiation to burnups higher than that possible with natural uranium. To provide a basis for the economic use of these fuel cycles, a program is underway to develop and demonstrate a CANDU fuel bundle capable of both higher burnups and greater operating margins. This new bundle design is being developed jointly by AECL and KAERI, and uses smaller-diameter fuel elements in the outer ring of a 43-element bundle to reduce the maximum element ratings in a CANDU fuel bundle by 20% compared to the 37-element bundle currently in use. This allows operation to burnups greater than 21 MWd/KgU. A combination of this lower peak-element rating, plus development work underway at AECL to enhance the thermalhydraulic characteristics of the bundle (including both critical heat flux and bundle pressure drop), provides a greater operating margin for the bundle. This new bundle design is called CANFLEX, and the program for its development in Canada and Korea is described in this paper. (author). 19 refs, 5 figs

  15. Fuel bundle loss of cooling inside the fuelling machine at CANDU6 PHWR

    This article describes the that loss of forced circulation cooling flow of induce spent fuel bundle loss of cooling and fission product releasing, analyzes the effect of reactor building and environment due to the fuel bundle rupturing, discusses the countermeasure to deal with the event of loss of cooling of spent fuel bundle. (authors)

  16. Uranium's transformation from mineral to fuel bundles

    Uranium undergoes chemical transformation phases before it can be used in the nuclear power plant. In first phase uranium is transformed from mineral to yellow cake, in which uranium is in the form of U3O8. After that comes conversion (U3O8-UF6) and enrichment (0.7%-3% U-235). Finally, uranium is converted in fuel fabrication to uranium dioxide, UO2, and fuel pellets are made

  17. CAT reconstruction and potting comparison of a LMFBR fuel bundle

    A standard Liquid Metal Fast Breeder Reactor (LMFBR) subassembly used in the Experimental Breeder Reactor II (EBR-II) was investigated, by remote techniques, for fuel bundle distortion by both nondestructive and destructive methods, and the results from both methods were compared. The non-destructive method employed neutron tomography to reconstruct the locations of fuel elements through the use of a maximum entropy reconstruction algorithm known as MENT. The destructive method consisted of ''potting'' (a technique that embeds and permanently fixes the fuel elements in a solid matrix) the subassembly, and then cutting and polishing the individual sections. The comparison indicated that the tomography reconstruction provided good results in describing the bundle geometry and spacer-wire locations, with the overall resolution being on the order of a spacer-wire diameter. A dimensional consistency check indicated that the element and spacer-wire dimensions were accurately reproduced in the reconstruction

  18. Effect of Candu Fuel Bundle Modeling on Sever Accident Analysis

    Dupleac, D.; Prisecaru, I. [Power Plant Engineering Faculty, Politehnica University, 313 Splaiul Independentei, 060042, sect. 6, Bucharest (Romania); Mladin, M. [Institute for Nuclear Research, Pitesti-Mioveni, 115400 (Romania)

    2009-06-15

    In a Candu 6 nuclear power reactor fuel bundles are located in horizontal Zircaloy pressure tubes through which the heavy-water coolant flows. Each pressure tube is surrounded by a concentric calandria tube. Outside the calandria tubes is the heavy-water moderator contained in the calandria itself. The moderator is maintained at a temperature of 70 deg. C by a separate cooling circuit. The moderator surrounding the calandria tubes provides a potential heat sink following a loss of core heat removal. The calandria vessel is in turn contained within a shield tank (or reactor vault), which provides biological shielding during normal operation and maintenance. It is a large concrete tank filled with ordinary water. During normal operation, about 0.4% of the core's thermal output is deposited in the shield tank and end shields, through heat transfer from the calandria structure and fission heating. In a severe accident scenario, the shield tank could provide an external calandria vessel cooling which can be maintained until the shield tank water level drops below the debris level. The Candu system design has specific features which are important to severe accidents progression and requires selective consideration of models, methods and techniques of severe accident evaluation. Moreover, it should be noted that the mechanistic models for severe accident in Candu system are largely less well validated and as the result the level of uncertainty remains high in many instances. Unlike the light water reactors, for which are several developed computer codes to analyze severe accidents, for Candu severe accidents analysis two codes were developed: MAAP4-Candu and ISAAC. However, both codes started by using MAAP4/PWR as reference code and implemented Candu 6 specific models. Thus, these two codes had many common features. Recently, a joint project involving Romanian nuclear organizations and coordinated by Politehnica University of Bucharest has been started. The purpose

  19. Improving the useful life of a 37-element fuel bundle

    Preliminary results indicate that CANDU burnup using 37-element fuel bundle with a slight enrichment can improve the useful life in the core. A slight enrichment in this study is increasing U-235 from 0.72 to 0.9 mass percent. A parametric study on criticality using Atomic Energy of Canada Limited’s WIMSAECL 3.1 and the Monte Carlo code, MCNP 5, developed by Los Alamos National Laboratory, is presented in this paper. (author)

  20. Fuel element bundle shears with dust extraction when cutting

    To prevent deposits of dust when cutting in this very inaccessible area of the fuel element bundle shears, a grating is fitted, which is connected via extraction devices (a collecting funnel and extraction duct) to the downward shaft carrying flushing air for the pipe pieces cut off. The measures taken make it possible to remove dust during cutting by the joint action of flushing air and gravity. (orig./HP)

  1. CANDU-6 fuel bundle fabrication and advanced fuels development in China

    In recent years, China North Nuclear Fuel Corporation (CNNFC) has introduced several modifications to the manufacturing processes and the production line equipment. This has been beneficial in achieving a very high level of quality in the production of fuel bundles. Since 2008 CNNFC has participated in a multi party project with the goal of developing advanced fuels for use in CANDU reactors. Other project team members include the Nuclear Power Institute of China (NPIC), Third Qinshan Nuclear Power Company (TQNPC) and Atomic Energy of Canada Ltd (AECL). This paper will present the improvements developed during the manufacture of natural fuel bundles and advanced fuels. (author)

  2. Hydraulic reinforcement of channel at lower tie-plate in BWR fuel bundle

    This patent describes an apparatus in a fuel bundle for confining fuel rods for the generation of steam in a steam water mixture passing interior of the fuel bundle. The fuel bundle includes: a lower tie-plate for supporting the fuel rods and permitting flow from the lower exterior portion of the fuel bundle into the interior portion of the fuel bundle; a plurality of fuel rods. The fuel rods supported on the lower tie-plate extending upwardly to and towards the upper portion of the fuel bundle for the generation of steam in a passing steam and water mixture interior of the fuel bundle; an upper tie-plate for maintaining the fuel rods in side-by-side relation and permitting a threaded connection between a plurality of the fuel rods with the threaded connection being at the upper and lower tie-plate. The upper tie-plate permitting escape of a steam water mixture from the top of the fuel bundle; a fuel bundle channel; and a labyrinth seal configured in the lower tie-plate

  3. Investigations on flow induced vibration of simulated CANDU fuel bundles in a pipe

    In this paper, vibration of a two-bundle string consisting of simulated CANDU fuel bundles subjected to turbulent liquid flow is investigated through numerical simulations and experiments. Large eddy simulation is used to solve the three-dimensional turbulent flow surrounding the fuel bundles for determining fluid excitations. The CFD model includes pipe flow, flow through the inlet fuel bundle along with its two endplates, half of the second bundle and its upstream endplate. The fluid excitation obtained from the fluid model is subsequently fed into a fuel bundle vibration code written in FORTRAN. Fluid structure interaction terms for the fuel elements are approximated using the slender body theory. Simulation results are compared to measurements conducted on the simulated fuel bundles in a testing hydraulic loop. (author)

  4. HLM fuel pin bundle experiments in the CIRCE pool facility

    Martelli, Daniele, E-mail: daniele.martelli@ing.unipi.it [University of Pisa, Department of Civil and Industrial Engineering, Pisa (Italy); Forgione, Nicola [University of Pisa, Department of Civil and Industrial Engineering, Pisa (Italy); Di Piazza, Ivan; Tarantino, Mariano [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone (Italy)

    2015-10-15

    Highlights: • The experimental results represent the first set of values for LBE pool facility. • Heat transfer is investigated for a 37-pin electrical bundle cooled by LBE. • Experimental data are presented together with a detailed error analysis. • Nu is computed as a function of the Pe and compared with correlations. • Experimental Nu is about 25% lower than Nu derived from correlations. - Abstract: Since Lead-cooled Fast Reactors (LFR) have been conceptualized in the frame of GEN IV International Forum (GIF), great interest has focused on the development and testing of new technologies related to HLM nuclear reactors. In this frame the Integral Circulation Experiment (ICE) test section has been installed into the CIRCE pool facility and suitable experiments have been carried out aiming to fully investigate the heat transfer phenomena in grid spaced fuel pin bundles providing experimental data in support of European fast reactor development. In particular, the fuel pin bundle simulator (FPS) cooled by lead bismuth eutectic (LBE), has been conceived with a thermal power of about 1 MW and a uniform linear power up to 25 kW/m, relevant values for a LFR. It consists of 37 fuel pins (electrically simulated) placed on a hexagonal lattice with a pitch to diameter ratio of 1.8. The FPS was deeply instrumented by several thermocouples. In particular, two sections of the FPS were instrumented in order to evaluate the heat transfer coefficient along the bundle as well as the cladding temperature in different ranks of sub-channels. Nusselt number in the central sub-channel was therefore calculated as a function of the Peclet number and the obtained results were compared to Nusselt numbers obtained from convective heat transfer correlations available in literature on Heavy Liquid Metals (HLM). Results reported in the present work, represent the first set of experimental data concerning fuel pin bundle behaviour in a heavy liquid metal pool, both in forced and

  5. Post-irradiation examination of the 37M fuel bundle at Chalk River Laboratories (AECL)

    Armstrong, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Daniels, T. [Ontario Power Generation, Pickering, Ontario (Canada); Montin, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2014-03-15

    The modified (-element (37M) fuel bundle was designed by Ontario Power Generation (OPG) to improve Critical Heat Flux (CHF) performance in ageing pressure tubes. A modification of the conventional 37-element fuel bundle design, the 37M fuel bundle allows more coolant flow through the interior sub-channels by way of a smaller central element. A demonstration irradiation (DI) of thirty-two fuel bundles was completed in 2011 at OPG's Darlington Nuclear Generating Station to confirm the suitability of the 37M fuel bundles for full core implementation. In support of the DI, fuel elements were examined in the Chalk River Laboratories Hot Cells. Inspection activities included: Bundle and element visual examination; Bundle and element dimensional measurements; Verification of bundle and element integrity; and Internal Gas Volume Measurements. The inspection results for 37M were comparable to that of conventional 37-element CANDU fuel. Fuel performance parameters of the 37M DI fuel bundle and fuel elements were within the range observed for similarly operated conventional 37-element CANDU fuel. Based on these Post Irradiation Examination (PIE) results, 37M fuel performed satisfactorily. (author)

  6. An assessment of thermal behavior of the DUPIC fuel bundle by subchannel analysis

    Thermal behavior of the standard DUPIC fuel has been assessed. The DUPIC fuel bundle has been modeled for a subchannel analysis using the ASSERT-IV code which was developed by AECL. From the calculated mixture enthalpy, equilibrium quality and void fraction distributions of the DUPIC fuel bundle, 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. Based upon the subchannel modeling used in this study, the location of minimum CHFR in the DUPIC fuel bundle has been found to be very similar to that of the standard fuel. 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 was found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. Since the transverse interchange model between subchannels is important for the behavior of these variables, it is needed to put more effort in validating the transverse interchange model. For the purpose of investigating influence of thermal-hydraulic parameter variations of the DUPIC fuel bundle, four different values of the channel flow rates were used in the subchannel analysis. The effect of the channel flow reduction on thermal-hydraulic parameters have been presented. This study shows that the subchannel analysis is very useful in assessing thermal behavior of the fuel bundles in CANDU reactors. (author). 12 refs., 3 tabs., 17 figs

  7. Application of Sipping and Visual Inspection Systems for the Evaluation of Spent Fuel Bundle Integrity

    When CANDU reactor has defective fuel bundle during its operation, then the defective fuel bundle should be discharged by 2(two) fuel bundles at a time from the corresponding fuel channel until the failed fuel bundle is found. Existing fuel failure detection system GFP(Gaseous Fission Product) & DN(Delayed Neutron) Monitoring System can’t exactly distinguish fuel elements failure from each fuel bundle. Because of fuelling machine mechanism and discharge procedure, always two fuel bundles at a time are being inspected. In case visual inspection is available for inspecting fuel elements and suppose that there are no defects and damaged marks on the surface of outer fuel elements, 2(two) defective fuel bundles should be canned and kept in the separate region of spent fuel storage pool. Therefore, the purpose of this study was to develop a system which is capable of inspecting whether each fuel bundle is failed or not. KNF (KEPCO Nuclear Fuel Co. Ltd) developed two evaluation systems to investigate the integrity of CANDU spent fuel bundle. The first one is a sipping system that detects fission gases leaked from fuel element. The second one is a visual inspection system with radiation resistant underwater camera and remotely controlled devices. The sipping technology enables to analyze the leakage of fission products not only in gaseous state but also liquid state. The performance of developed systems was successfully demonstrated at Wolsong power plant this year. This paper describes the results of the development of the failed fuel detection technology and its application. (author)

  8. Spent fuel bundle counter sequence error manual - DARLINGTON NGS

    The Spent Fuel Bundle Counter (SFBC) is used to count the number and type of spent fuel transfers that occur into or out of controlled areas at CANDU reactor sites. However if the transfers are executed in a non-standard manner or the SFBC is malfunctioning, the transfers are recorded as sequence errors. Each sequence error message typically contains adequate information to determine the cause of the message. This manual provides a guide to interpret the various sequence error messages that can occur and suggests probable cause or causes of the sequence errors. Each likely sequence error is presented on a 'card' in Appendix A. Note that it would be impractical to generate a sequence error card file with entries for all possible combinations of faults. Therefore the card file contains sequences with only one fault at a time. Some exceptions have been included however where experience has indicated that several faults can occur simultaneously

  9. A model for fuel rod and tie rod elongations in boiling water reactor fuel bundles

    A structural model is developed to determine the relative axial displacements of the spring held fuel rods to the tie rods in Boiling Water Reactor fuel bundles. An irradiation dependent relaxation model, which considers a two stage relaxation process dependent upon the fast fluence is used for the compression springs. The changes in spring compression resulting from the change in the length of the zircaloy fuel cladding due to irradiation enhanced anisotropic creep and growth is also considered in determining the time dependent variation of the spring forces. The time dependence of the average linear heat generation rates and their axial distributions is taken into account in determining the fuel cladding temperatures and fast fluxes for the various fuel rod locations within each of the BWR fuel bundles whose relative displacements were measured and used in this verification study. (orig.)

  10. Studies of a larger fuel bundle for the ABWR improved evolutionary reactor

    Studies for an Improved Evolutionary Reactor (IER) based on the Advanced Boiling Water Reactor (ABWR) were initiated in 1990. The author summarizes the current status of the core and fuel design. A core and fuel design based on a BWR K-lattice fuel bundle with a pitch larger than the conventional BWR fuel bundle pitch is under investigation. The core and fuel design has potential for improved core design flexibility and improved reactor transient response. Furthermore, the large fuel bundle, coupled with a functional control rod layout, can achieve improvement of operation and maintenance, as well as improvement of overall plant economy

  11. LVRF fuel bundle manufacture for Bruce - project update

    In response to the Power Uprate program at Bruce Power, Zircatec has committed to introduce, by Spring 2006 a new manufacturing line for the production of 43 element Bruce LVRF bundles containing Slightly Enriched Uranium (SEU) with a centre pin of blended dysprosia/urania (BDU). This is a new fuel design and is the first change in fuel design since the introduction of the current 37 element fuel over 20 years ago. Introduction of this new line has involved the introduction of significant changes to an environment that is not used to rapid changes with significant impact. At ZPI we have been able to build on our innovative capabilities in new fuel manufacturing, the strength and experience of our core team, and on our prevailing management philosophy of 'support the doer'. The presentation will discuss some of the novel aspects of this fuel introduction and the mix of innovative and classical project management methods that are being used to ensure that project deliverables are being met. Supporting presentations will highlight some of the issues in more detail. (author)

  12. Optimized critical power in a fuel bundle with part length rods

    Johansson, E.B.; Matzner, B.; Dix, G.E.; Wolters, R.A. Jr.; Reese, A.P.

    1993-07-20

    In a boiling water reactor having discrete bundles of fuel rods confined within channel enclosed fuel assemblies wherein the fuel bundle includes: a plurality of fuel rods for placement within said channel, each fuel rod containing fissile material for producing nuclear reaction; a lower tie plate for supporting the bundle of fuel rods within said channel, the lower tie plate joining the bottom of the channel to close the bottom end of the channel, the lower tie plate providing defined apertures for the inflow of water coolant in the channel between the fuel rods for generation of steam; the plurality of fuel rods extending from the lower tie plate wherein a single phase region of the water in the bundle is defined to an upward portion of the bundle wherein an annular flow regime of the water and steam in the bundle is defined during nuclear steam generating reaction; an upper tie plate for supporting the upper end of the bundle of fuel rods, the upper tie plate joining the top of the channel, the upper tie plate providing apertures for the outflow of water and generated steam in the channel; spacers intermediate the upper and lower tie plates at preselected elevations along the fuel rods for maintaining the fuel rods in spaced apart location along the length of the fuel assembly including a first group of spacers in thelower region of the fuel bundle and a second group of spacers in the upper annular flow regime of the fuel bundle; a plurality of the fuel rods being part length extending from thelower tie plate towards the upper tie plate, the partial length fuel rods terminating at ends within the upper region of the fuel bundle before reaching the upper tie plate and causing deceased pressure drop in said annular flow regime of said fuel bundle during said nuclear steam generating reaction; the improvement to said bundle comprising: means in the annular flow regime of the fuel bundle for restoring at least some of the decreased pressure drop.

  13. Optimized critical power in a fuel bundle with part length rods

    In a boiling water reactor having discrete bundles of fuel rods confined within channel enclosed fuel assemblies wherein the fuel bundle includes: a plurality of fuel rods for placement within said channel, each fuel rod containing fissile material for producing nuclear reaction; a lower tie plate for supporting the bundle of fuel rods within said channel, the lower tie plate joining the bottom of the channel to close the bottom end of the channel, the lower tie plate providing defined apertures for the inflow of water coolant in the channel between the fuel rods for generation of steam; the plurality of fuel rods extending from the lower tie plate wherein a single phase region of the water in the bundle is defined to an upward portion of the bundle wherein an annular flow regime of the water and steam in the bundle is defined during nuclear steam generating reaction; an upper tie plate for supporting the upper end of the bundle of fuel rods, the upper tie plate joining the top of the channel, the upper tie plate providing apertures for the outflow of water and generated steam in the channel; spacers intermediate the upper and lower tie plates at preselected elevations along the fuel rods for maintaining the fuel rods in spaced apart location along the length of the fuel assembly including a first group of spacers in thelower region of the fuel bundle and a second group of spacers in the upper annular flow regime of the fuel bundle; a plurality of the fuel rods being part length extending from thelower tie plate towards the upper tie plate, the partial length fuel rods terminating at ends within the upper region of the fuel bundle before reaching the upper tie plate and causing deceased pressure drop in said annular flow regime of said fuel bundle during said nuclear steam generating reaction; the improvement to said bundle comprising: means in the annular flow regime of the fuel bundle for restoring at least some of the decreased pressure drop

  14. Modelling of fuel bundle deformation at high temperatures: requirements, models and steps for consideration

    To model thermal mechanical bundle deformation behaviour under high temperature conditions, several factors need to be considered. These are the sources of loads, deformation mechanisms, interactions within bundle components, bundle and pressure tube (PT) interaction, and boundary constraints on the fuel bundles under in-reactor conditions. This paper describes the modelling of the following three processes: Bundle slumping due to high temperature creep-sag of individual elements and endplates; Differential element expansion and fuel element bowing; and, Bundle distortion under axial loads. To model these processes, a number of key mechanisms for bundle deformation must be considered, which include: 1) Interaction of fuel elements in a bundle with their neighbours, 2) Endplate deformation, 3) Fuel elements lateral deformation under various loads and mechanisms, 4) Interaction within a fuel element, 5) Material property change at high temperatures, 6) Transient response of a bundle, and 7) Bundle configuration change. This paper summarises the new models needed for the mechanistic modelling of the key mechanisms mentioned above and provides an example to show how an endplate plasticity model is developed with results. (author)

  15. Fuel bundle examination techniques for the Phebus fission product test

    The paper develops the non-destructive examinations, with a special emphasis on transmission tomography, performed in the Phebus facility, using a linear accelerator associated with a line scan camera based on PCD components. This particular technique enabled the high level of penetration to be obtained, necessary for this high density application. Spatial resolution is not far from the theoretical limit and the density resolution is often adequate. This technique permitted: 1) to define beforehand the cuts on a precise basis, avoiding a long step-by-step choice as in previous in-pile tests; 2) to determine, at an early stage, mass balance, material relocations (in association with axial gamma spectrometry), and FP distribution, as an input into re-calculations of the bundle events. However, classical cuttings, periscopic visual examinations, macrographies, micrographies and EPMA analyses remain essential to give oxidation levels (in the less degraded zones), phase aspect and composition, to distinguish between materials of identical density, and, if possible, to estimate temperatures. Oxidation resistance of sensors (thermocouples or ultrasonic thermometers) is also traced. The EPMA gives access to the molten material chemical analyses, especially in the molten fuel blockage area. The first results show that an important part of the fuel bundle melted (which was one of the objectives of this test) and that the degradation level is close to TIMI-2 with a molten plug under a cavity surrounded by an uranium-rich crust. In lower and upper areas fuel rods are less damaged. Complementaries between these examination techniques and between international teams involved will be major advantages in the Phebus FPT0 test comprehension. 3 refs, 9 figs

  16. Designing for maintainability: use of operating experience and feedback to improve performance of the ACR-1000; detailing the specific case of power operated valves

    The AECL, ACR-1000 is being designed to achieve high lifetime capacity factor, low unplanned forced outages and short planned outages once per 3 years. In order to achieve this target extensive use is made of operating experience and industry feedback to improve the plant design. There is also a target to operate and maintain the plant with less staff than current CANDU power plants. The design will accommodate improvements in staff productivity. While the focus is set on the targets above there is also one more objective, to contain the capital cost of the new plants. AECL and partners are designing the ACR-1000 plant using a number of initiatives that are client driven [e.g. through COG]. This paper outlines the use of industry feedback in general with specific details for Power Operated Valves (POVs). The nuclear industry has promoted that the basis for good operation and maintenance is best achieved by building in improvements in the initial design. AECL has endorsed this philosophy and feedback is central to the design of the ACR-1000. (author)

  17. CFD and DNS methodologies development for fuel bundle simulations

    Development and application of Computational Fluid Dynamics (CFD) and Direct Numerical Simulation (DNS) approaches to the simulation of coolant flow inside nuclear fuel bundles are presented, focusing on the advantages and limitations of the different methodologies and on their synergetic potential. High Reynolds number flow cases are analyzed with the adoption of an improved anisotropic turbulence modeling, which adopts a non-linear stress strain correlation and an improved near wall treatment. The capability of the model of predicting the coolant flow distribution inside the bundles is shown and discussed on the base of comparison with experimental data for a variety of geometrical and Reynolds number conditions. In particular wall shear stresses, velocity, and secondary flow distributions comparisons are shown. Moreover, DNS computations are performed adopting an algorithm based on the finite difference method, extended to boundary fitted coordinate systems in order to efficiently concentrate grids near the distorted wall boundaries. The validity and significance of the results is discussed underlying the importance of the insights into the turbulence structure. The calculations are further extended to higher Reynolds numbers, which cannot in general be treated with DNS approach, renouncing to the estimation of the higher-order moments, but limited to the evaluation of the averaged velocity profiles, turbulence intensities and Reynolds stresses. (authors)

  18. The Conflux Fuel bundle: An Economic and Pragmatic Route to the use of Advanced Fuel Cycles in CANDU Reactors

    The CANFLEX1 bundle is being developed jointly by AECL and KAERI as a vehicle for introducing the use of enrichment and advanced fuel cycles in CANDU2 reactors. The bundle design uses smaller diameter fuel elements in the outer ring of a 43-element bundle to reduce the maximum element ratings in a CANDU fuel bundle by 20% compared to the 37-element bundle currently in use. This facilitates burnups of greater than 21,000 MW d/TAU to optimize the economic benefit available from the use of enrichment and advanced fuel cycles. A combination of this lower fuel rating, plus development work underway at Aecl to enhance the thermalhydraulic characteristics of the bundle (including both CHF3 and bundle. This provides extra flexibility in the fuel management procedures required for fuel bundles with higher fissile contents. The different bundle geometry requires flow tests to demonstrate acceptable vibration and fretting behavior of the Conflux bundle. A program to undertake the necessary range of flow tests has started at KAERI, involving the fabrication of the required bundles, and setting up for the actual tests. A program to study the fuel management requirements for slightly enriched (0.9 wt % 235 in total U) Conflux fuel has been undertaken by both Aecl and KAERI staff, and further work has started for higher enrichments. Irradiation testing of the Conflux bundle started in the NUR reactor in 1989, and a second irradiation test is due to start shortly. This paper describes the program, and reviews the status of key parts of the program

  19. Laser dismantling of PHWR spent fuel bundles and decladding of fuel pins in the highly radioactive hot cells

    Full text: For reprocessing of PHWR fuel, fuel bundles are at present chopped mechanically into small pieces of pins using high tonnage mechanical press before dissolution. The existing method of bundle dismantling is purely mechanical using very high force for chopping. A laser based automated bundle dismantling system is developed. In the system, end-plates of bundle, which holds the fuel pins together, are cut using Nd-YAG laser to separate the bundles into pins. In addition to pin separation, the pins are to be chopped into small pieces using a small mechanical chopper. Since the spent fuel is highly radioactive, all these operations are performed remotely in hot cells. Post irradiation examination also requires dismantling of bundles into pins so that they can select the pins for the further examinations. In both these applications laser dismantling remains the most. important step and this system has been developed and tested. This paper describes the experience gained during the development efforts

  20. Subchannel analysis of CANDU 37-element fuel bundles

    The subchannel analysis codes COBRA-IV and ASSERT-4 have been used to predict the mass and enthalpy imbalance within a CANDU 37-element fuel channel under various system conditions. The objective of this study was to assess the various capabilities of the ASSERT code and highlight areas where further validation or development may be needed. The investigation indicated that the ASSERT code has all the basic models required to accurately predict the flow and enthalpy imbalance for complex rod bundles. The study also showed that the code modelling of void drift and diffusion requires refinement to some coefficients and that further validation is needed at high flow rate and high void fraction conditions, where ASSERT and COBRA are shown to predict significantly different trends. The results of a recent refinement of ASSERT modelling are also discussed

  1. Cap assembly for a bundled tube fuel injector

    LeBegue, Jeffrey Scott; Melton, Patrick Benedict; Westmoreland, III, James Harold; Flanagan, James Scott

    2016-04-26

    A cap assembly for a bundled tube fuel injector includes an impingement plate and an aft plate that is disposed downstream from the impingement plate. The aft plate includes a forward side that is axially separated from an aft side. A tube passage extends through the impingement plate and the aft plate. A tube sleeve extends through the impingement plate within the tube passage towards the aft plate. The tube sleeve includes a flange at a forward end and an aft end that is axially separated from the forward end. A retention plate is positioned upstream from the impingement plate. A spring is disposed between the retention plate and the flange. The spring provides a force so as to maintain contact between at least a portion of the aft end of the tube sleeve and the forward side of the aft plate.

  2. MENT reconstruction and potting comparison of a LMFBR fuel bundle

    Since the advent of computer-assisted-tomography (CAT), the CAT techniques have been rapidly expanded to the nuclear industry. A number of investigators have applied these techniques to reconstruct the fuel bundle configuration inside a subassembly with various degrees of resolution; however, there has been little data available on the accuracy of these reconstructions, and no comparisons have been made with the internal structure of actual irradiated subassemblies. Some efforts have utilized pretest mock-ups to calibrate the CAT algorithms, but the resulting mock-up configurations do not necessarily represent an actual subassembly, so an exact comparison has been lacking. The purpose of this paper is to present the results of a comparison between a CAT reconstruction of an irradiated subassembly and the destructive examination of the same subassembly

  3. Development of the cooling technology on TRU fuel pin bundle during fuel fabrication process (4). Steady state cooling test of full mock up fuel pin bundle

    The development of the fast reactor cycle is being preceded in Japan to utilize plutonium and trans-uranium materials which come from the simplified PUREX reprocessing. But the TRU fuel bundle generates heat due to fission of TRU during the fabrication process of the wire wrapped Fast Breeder Reactor (FBR) fuel pin bundle. Then it is a big issue to develop an efficient cooling system for the horizontally laid bundle and to clarify its thermal behavior. Then in this paper the steady state full mock up test results are described. Inlet air velocity and heat generation rate were varied in the tests as the parameter. Then it is ascertained that the fuel can be cooled under the 473 K which is the criterion for the steady state cooling of this study to keep cladding soundness. The temperature and velocity fields of the bundle upper side were also measured by moving thermocouples to vertical and horizontal directions, by the infrared thermometer and by PIV (Particle Image Velocimetry). Then the temperature and velocity fields at outlet region are clarified. (author)

  4. Nanocrystal and noble gas tagging for monitoring defective CANDU fuel bundles

    The purpose of this paper is to discuss two possible defective fuel bundle tagging techniques that have been suggested for CANDU-6 nuclear reactors. The general design of a CANDU-6 reactor and fuel bundle is reviewed. Nanocrystal tagging is introduced. A current production method for CdTe nanocrystals and future experimental goals are outlined and noble gas tagging is reviewed. Considerations for the future implementation of these tagging methods for fuel in a CANDU-6 reactor is also discussed. (author)

  5. Automation in inspection of PHWR fuel elements & bundles at Nuclear Fuel Complex

    Nuclear Fuel Complex (NFC), Hyderabad, a constituent of Department of Atomic Energy, India manufactures fuel for all Indian nuclear power reactors. Currently NFC manufactures both 19 element & 37 element bundles for catering to the requirement of 220 MWe & 540 MWe PHWRs. In order to meet the growing needs for the Nuclear Fuel, NFC engaged in expansion of the production facilities. This calls for enhanced throughput at various inspection stages keeping in tandem with the production & for achieving this objective, NFC has chosen automation. This paper deals with automation of the inspection line at NFC. (author)

  6. Upon local blockage formations in LMFBR fuel rod bundles with wire-wrapped spacers

    A theoretical and experimental study, to improve understanding of local particle depositions in a wire-wrapped LMFBR fuel bundle, has been performed. Theoretical considerations show, that a preferentially axial process of particle depositions occurs. The experiments confirm this and clarify that the blockages arise near the particle source and settle at the spatially arranged minimum gaps in the bundle. The results suggest that, considering flow reduction, cooling and DND-detection, such fuel particle blockages are less dangerous. With reference to these safety-relevant factors, wire-wrapped LMFBR fuel bundles seem to gain advantages compared to the grid design. (orig.)

  7. Testing and implementation program for the modified Darlington 37-element fuel bundle

    To mitigate the effects of reactor ageing, a design modification to the 37-element fuel is proposed in which the diameter of the centre element will be reduced to 11.5 mm from 13.1 mm. The testing and implementation phase for the 37-element fuel bundle modification is discussed in this paper. The initial plan for testing is to perform a set of out-reactor tests to assess the endurance, acoustic response and cross-flow behaviour of the revised fuel bundle design. The initial schedule outlines activities that will enable OPG to implement full core fuelling of the modified bundle within the next three to four years. (author)

  8. Full-Scale Irradiation Test of Hanaro U3Si Fuel Using Lead Bundle

    To verify the irradiation performances of HANARO fuel at a nominal power of 30 MW, a lead bundle was first loaded into the HANARO core after increasing the reactor power to the full power. The lead bundle is an actual fuel assembly with 18 fuel rods that was fabricated using an atomized manufacturing procedure. The lead bundle was irradiated during 188 operation days at full power in the HANARO core, and discharged after about 60 at% average and 75 at% peak burn-ups. The maximum linear power of the lead bundle was 98kW/m. Detailed non-destructive and destructive post-irradiation tests were performed. The measured results were analyzed and compared with the existing experimental data and the design criteria for the HANARO fuel. It was confirmed that the HANARO fuel has maintained proper in-pile performances and integrity during the nominal power operation and satisfies all the design requirements related to the irradiation performances. (author)

  9. Manufacturing of 37-element fuel bundles for PHWR 540 - new approach

    Nuclear Fuel Complex (NFC), established in early seventies, is a major industrial unit of Department of Atomic Energy. NFC is responsible for the supply of fuel bundles to all the 220 MWe PHWRs presently in operation. For supplying fuel bundles for the forthcoming 540 MWe PHWRs, NEC is dovetailing 37-element fuel bundle manufacturing facilities in the existing plants. In tune with the philosophy of self-reliance, emphasis is given to technology upgradation, higher customer satisfaction and application of modern quality control techniques. With the experience gained over the years in manufacturing 19-element fuel bundles, NEC has introduced resistance welding of appendages on fuel tubes prior to loading of UO2 pellets, use of bio-degradable cleaning agents, simple diagnostic tools for checking the equipment condition, on line monitoring of variables, built-in process control methods and total productive maintenance concepts in the new manufacturing facility. Simple material handling systems have been contemplated for handling of the fuel bundles. This paper highlights the flow-sheet adopted for the process, design features of critical equipment and the methodology for fabricating the 37-element fuel bundles, 'RIGHT FIRST TIME'. (author)

  10. RU-43 a new uranium fuel bundle design for using in CANDU type reactors

    A unique feature of the CANDU reactor design is its ability to use alternative fuel cycles other than natural uranium (NU), without requiring major modifications to the basic reactor design. These alternative fuel cycles, which are known as advanced fuel cycles, utilize a variety of fissile materials, including Slightly Enriched Uranium (SEU) from enrichment facilities, and Recovered Uranium (RU) obtained from the reprocessing of the spent fuel of light-water reactors (LWR). A fissile content in the RU of 0.9 to 1.0 % makes it impossible for reuse in an LWR without re-enrichment, but CANDU reactors have a sufficient high neutron economy to use RU as fuel. RU from spent LWR fuel can be considered as a lower cost source of enrichment at the optimal enrichment level for CANDU fuel pellets. In Europe the feedstock of RU is approaching thousands tones and would provide sufficient fuel for hundreds CANDU-6 reactors years of operation. The use of RU fuel offers significant benefits to CANDU reactor operators. RU fuels improve fuel cycle economics by increasing the fuel burnup, which enables large cost reductions in fuel consumption and in spent fuel disposal. RU fuel offers enhanced operating margins that can be applied to increase reactor power. These benefits can be realized using existing fuel production technologies and practices, and with almost negligible changes to fuel receipt and handling procedures at the reactor. The application of RU fuel could be an important element in NPP Cernavoda from Romania. For this reason the Institute for Nuclear Research (INR), Pitesti has started a research programme aiming to develop a new fuel bundle RU-43 for extended burnup operation. The current version of the design is the result of a long process of analyses and improvements, in which successive preliminary design versions have been evaluated. The most relevant calculations performed on this fuel element design version are presented. Also, the stages of an experimental

  11. Analysis of the operational reliability of VVER-1000 fuel elements and bundles in a three-year fuel cycle

    At the Novo-Voronezh Nuclear Power Plant, the fifth VVER-1000 unit, which was operated at nominal power from February 1980, completed nine fuel cycles in July 1990. The first unit of the Kalinin Nuclear Power Plant has operated from April 1984; in October 1990 the sixth fuel loading was completed. To data these power units are operating in steady-state in three-year fuel cycles (from June 1986 and from September 1989, respectively). By the end of 1988, operational experience had been accumulated on 1407 fuel element bundles on the third to the sixth fuel loading at Kalinin and the fifth to the ninth at Novo-Voronezh, which are in the transient and steady-state regimes of a three-year cycle. Of the 561 fuel element bundles monitored for gamma radiation, 14 were designated as leaking, which was 2.5% of the total bundles or 0.008% of the total number of fuel elements. Thus, a high degree of reliability was attained with enriched fuel elements. Here the authors analyze the reliability of fuel element bundles in taking the VVER-1000s to a three-year fuel cycle, and also generalize and systematize information on the fundamental characteristics of a group of fuel element bundles in going to to steady-state conditions of the three-year fuel cycle

  12. Study of the end flux peaking for the Candu fuel bundle types by transport methods

    The region separating the Candu fuel in two adjoining bundles in a channel is called the end region. The end of the last pellet in the fuel stack adjacent to the end region is called the fuel end. In the end region of the bundle the thermal neutron flux is higher than at the axial mid-point, because the end region of the bundle is made up of very low neutron absorption material: coolant and Zircaloy-4. For accurate evaluation of fuel performance, it is important to have capability to calculate the three dimensional spatial flux distributions in the fuel bundle, including the end region. The work reported here had two objectives. First, calculation of the flux distributions (axial and radial) and the end flux peaking factors for some Candu fuel bundles. Second objective is a comparative analysis of the obtained results. The Candu fuel bundles considered in this paper are NU37 (Natural Uranium, 37 elements) and SEU43 (Slightly Enriched Uranium, 43 elements, with 1.1wt% enrichment). For realization of the proposed objectives, a methodology based on WIMS, PIJXYZ and LEGENTR codes is used in this paper. WIMS is a standard lattice-cell code, based on transport theory and it is used for producing fuel cell multigroup macroscopic cross sections. For obtaining the flux distribution in Candu fuel bundles it is used PIJXYZ and LEGENTR respectively codes. These codes are consistent with WIMS lattice-cell calculations and allow a good geometrical representation of the Candu bundle in three dimensions. PIJXYZ is a 3D integral transport code using the first collision probability method and it has been developed for Candu cell geometry. LEGENTR is a 3D SN transport code based on projectors technique and can be used for 3D cell and 3D core calculations. (author)

  13. The behaviour of Phenix fuel pin bundle under irradiation

    An entire Phenix sub-assembly has been mounted and sectioned after irradiation. The examination of cross-sections revealed the effects of mechanical interaction in the bundle (ovalisations and contacts between clads). According to analysis of the sodium channels, cooling of the pin bundle remained uniform. (author)

  14. CANFLEX fuel bundle strength tests during normal and abnormal refuelling procedure

    As one of verifications of the CANFLEX fuel bundle, the strength tests were performed by the double side-stop test for the simulation of normal fuel loading and the single side-stop test for the simulation of abnormal fuel loading. In both tests the load was applied by controlling the flow to obtain a desired pressure drop across the whole fuel string resulting in a specified hydraulic drag force on the test bundle. The test rig conditions for each test were 120 .deg. C and 11.2 MPa for 15 minutes. The test bundles against the side-stop simulators were measured and inspected carefully after the tests according to the measurement procedures. The inspection results showed the test bundles were intact and met the acceptance criteria

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

    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

  16. Design verification of the CANFLEX fuel bundle - quality assurance requirements for mechanical flow testing

    As part of the design verification program for the new fuel bundle, a series of out-reactor tests was conducted on the CANFLEX 43-element fuel bundle design. These tests simulated current CANDU 6 reactor normal operating conditions of flow, temperature and pressure. This paper describes the Quality Assurance (QA) Program implemented for the tests that were run at the testing laboratories of Atomic Energy of Canada Limited (AECL) and Korea Atomic energy Research Institute (KAERI). (author)

  17. The effects of bearing-pad height on the critical heat flux of CANFLEX fuel bundle

    In CANDU-6 fuel channel, the geometrical eccentricity exists between fuel bundle and horizontal pressure tube. Based on the water CHF(critical heat flux) tests of the full-scale CANFLEX(CANDU Flexible) bundle string with the current bearing-pads of 1.4mm height, it was found that the increase of bypassing flow decreased significantly the CHF of fuel bundle with increasing the creep rate of pressure tube. So, the additional improvement of heat transfer performance is anticipated by increasing the hight of bearing-pads(about 0.3 mm) and reducing the eccentricity of fuel bundle. This paper presented the effects of bearing-pad height on the CHF by examining the water CHF test data of CANFLEX fuel strings equipped with 1.7 mm and 1.8 mm high bearing-pads. It also showed the data trends of the boiling-length-averaged CHF with respect to the test system flow parameters and local flow conditions. The high bearing-pad bundle is increased in dryout power by 7 to 10%, compared to the current CANFLEX fuel bundle

  18. Analysis of fuel handling system for fuel bundle safety during station blackout in 500 MWe PHWR unit of India

    Situations of Station Blackout (SBO) i.e. postulated concurrent unavailability of Class Ill and Class IV power, could arise for a long period, while on-power refuelling or other fuel handling operations are in progress with the hot irradiated fuel bundles being anywhere in the system from the Reactor Building to the Spent Fuel Storage Bay. The cooling provisions for these fuel bundles are diverse and specific to the various stages of fuel handling operations and are either on Class Ill or on Class II power with particular requirements of instrument air. Therefore, during SBO, due to the limited availability of Class II power and instrument air, it becomes difficult to maintain cooling to these fuel bundles. However, some minimal cooling is essential, to ensure the safety of the bundles. As discussed in the paper, safety of these fuel bundles in the system and/or for those lying in the liner tube region of the reactor end fitting is ensured, during SBO, by resorting to passive means like 'stay-put', 'gravity- fill', 'D20- steaming' etc. for cooling the bundles. The paper also describes various consequences emanating from these cooling schemes. (author). 6 refs., 2 tabs., 8 figs

  19. Behavior of mixed-oxide fuel elements in a tight bundle under duty-cycle conditions

    The irradiation behavior of the TOB-10 fuel pins was comparable with that obtained in the single pin tests. There was no significant effect that could be directly attributed to tight bundle configuration. The postirradiation examination data provided information on the axial migration of cesium and its effect on cladding strain. Severe fuel/cladding chemical interaction (FCCI), which resulted in substantial cladding thinning and probably restricted venting of fission gas from the fuel column into the pin plena, apparently caused the earlier-than-expected cladding breaches in the D9-clad pins. No such severe FCCI was noted in the 316SS-clad pins. At the time of test termination, the overall cladding strain from creep and swelling was insufficient to cause bundle closure. Consequently, there would have been minimal pin bundle-duct interaction in the subassembly. Neither of the breaches appeared to be induced by pin bundle-duct interaction. (author)

  20. Development of Romanian SEU-43 fuel bundle for CANDU type reactors

    SEU-43 fuel bundle is a CANDU type fuel consisting of two element sizes, to reduce element ratings, while maintaining the same bundle power, and an uranium content very close to the uranium content of a standard 37-element bundle. In order to reduce the detrimental effects of the life limiting factors at extended burnup a set of solution have been adopted for fuel element design. As a part of the design verification program, experimental bundles have been fabricated and utilized in typical out of reactor tests conducted at the laboratories of INR, Pitesti. These tests simulated current CANDU-6 reactor normal operating conditions of flow, temperature and pressure. The results are in accordance with the specified acceptance criteria. (author)

  1. Overview of methods to increase dryout power in CANDU fuel bundles

    Groeneveld, D.C., E-mail: degroeneveld@gmail.com [Chalk River Laboratories, AECL, Chalk River (Canada); University of Ottawa, Department of Mechanical Engineering, Ottawa (Canada); Leung, L.K.H. [Chalk River Laboratories, AECL, Chalk River (Canada); Park, J.H. [Korean Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-06-15

    Highlights: • Small changes in bundle geometry can have noticeable effects on the bundle CHF. • Rod spacing devices can results in increases of over 200% in CHF. • CHF enhancement decays exponentially downstream from spacers. • CHF-enhancing bundle appendages also increase the post-CHF heat transfer. - Abstract: In CANDU reactors some degradation in the CCP (critical channel power, or power corresponding to the first occurrence of CHF in any fuel channel) will occur with time because of ageing effects such as pressure-tube diametral creep, increase in reactor inlet-header temperature, increased hydraulic resistance of feeders. To compensate for the ageing effects, various options for recovering the loss in CCP are described in this paper. They include: (i) increasing the bundle heated perimeter, (ii) optimizing the bundle configuration, (iii) optimizing core flow and flux distribution, (iv) reducing the bundle hydraulic resistance, (v) use of CHF-enhancing bundle appendages, (vi) more precise experimentation, and (vii) redefining CHF. The increase in CHF power has been quantified based on experiments on full-scale bundles and subchannel code predictions. The application of several of these CHF enhancement principles has been used in the development of the 43-rod CANFLEX bundle.

  2. Parametric study of thermo-mechanical behaviour of 19-element PHWR fuel bundle having AHWR fuel material

    AHWR Th-LEU of 4.3 weight % 235U enrichment is a fuel design option for its trial irradiation in Indian PHWRs. The important component of this option is the large enhancement in the average discharge burn-up from the core. A parametric study of the 19-element fuel bundle, with natural uranium currently is being used in all operating 220 MWe PHWRs, has been carried out for AHWR Th-LEU fuel material by computer code FUDA MOD2. The important fuel parameters such as fuel temperature, fission gas release, fuel swelling and sheath strain have been analyzed for required fuel performance. With Th-LEU, average discharge burnups of about 25,000 MW-d/TeHE can be achieved. The FUDA code (Fuel Design Analysis code) MOD2 version has been used in the fuel element analysis. The code takes into account the inter-dependence of different parameters like fuel pellet temperatures, pellet expansions, fuel-sheath gap heat transfer, sheath strain and stresses, fission gas release and gas pressures, fuel densification etc. Thermo-mechanical analysis of fuel element having AHWR material is carried out for the bundle power histories reaching up to design burn-up 40000 MWd/TeHE. The resultant parameters such as fuel temperature, sheath plastic strain and fission gas pressure for AHWR fuel element were compared with respective thermo-mechanical parameters for similar fuel bundle element with natural uranium as fuel material. (author)

  3. System for supporting a bundled tube fuel injector within a combustor

    LeBegue, Jeffrey Scott; Melton, Patrick Benedict; Westmoreland, III, James Harold; Flanagan, James Scott

    2016-06-21

    A combustor includes an end cover having an outer side and an inner side, an outer barrel having a forward end that is adjacent to the inner side of the end cover and an aft end that is axially spaced from the forward end. An inner barrel is at least partially disposed concentrically within the outer barrel and is fixedly connected to the outer barrel. A fluid conduit extends downstream from the end cover. A first bundled tube fuel injector segment is disposed concentrically within the inner barrel. The bundled tube fuel injector segment includes a fuel plenum that is in fluid communication with the fluid conduit and a plurality of parallel tubes that extend axially through the fuel plenum. The bundled tube fuel injector segment is fixedly connected to the inner barrel.

  4. Scratch preventing method of assembling nuclear fuel bundles, and the assembly

    This patent describes a method of assembling a bundle of nuclear fuel elements for service in a nuclear reactor. It comprises a group of fuel rod elements each arranged in a space apart, parallel array and thus secured by each element traversing through a series of spacing units positioned at intervals along the length of the grouped fuel rod elements and having openings for receiving the fuel rod elements traversing therethrough, consisting essentially of the steps of: providing a scratch resisting, temporary protective barrier consisting of a water soluble coating of sodium silicate covering the outer surface of the fuel rod elements, then assembling the fuel bundle by passing each of the fuel rod elements through the openings of a series of spacing units positioned at intervals to fit together an adjoined composite fuel bundle assembly of a spaced apart parallel array of the fuel rod elements secured with spacing units, and removing the scratch resisting, temporary protective barrier consisting of water soluble coating of sodium silicate from the assembled fuel bundle with hot water

  5. The fission gas release and gas pressure calculation for 19 element fuel bundle irradiated in KAPS-1 (Bundle no-56504)

    The thermo-mechanical analysis of fuel bundle is done using FUDA software program to calculate the fission gas release and pin pressure. The fission gas release analysis was done for the average fuel dimensions. In addition, a parametric study was also performed by varying the different parameters within their specified tolerances. The thermal conductivity calculation in the present analysis accounts for the density changes and temperature variation. The feed back of gap conductance change due to fission gas accumulation in pellet clad gap is considered in fuel temperature calculations. The present paper discusses the inputs to the FUDA, mathematical model used in calculation of fission gas release and results of gas release from the FUDA runs for the above discussed analysis. (author)

  6. CFD study on coolant mixing in VVER-440 fuel rod bundles and fuel assembly heads

    A CFD model of VVER-440 fuel assembly heads was developed based on the technical documentation of a full-scale test facility built in the Kurchatov Institute, Russia. Steady-state and transient calculations were performed to validate the model with a measurement set. Effects of the spatial resolution, turbulence models, difference schemes and different inlet boundary conditions were investigated. Inlet boundary conditions were determined with both the COBRA subchannel code and a fuel rod bundle CFD model that was built for this special purpose. The results were compared against experimental data. The sensitivity studies showed that a grid of about 8 million cells, high resolution scheme and BSL Reynolds stress model are suitable sets to provide accurate prediction for the signal of the in-core thermocouple. The best prediction was achieved with transient calculation using inlet boundary conditions generated with the CFD fuel rod bundle model. The results indicated that the coolant mixing is intensive but not perfect in the assembly head. Besides, the significant role of the outflow from the central tube was also proven. The transient runs revealed relatively large temperature fluctuations near the in-core thermocouple housing.

  7. Research reactor fuel bundle design review by means of hydrodynamic testing

    During the design steps of a fuel bundle for a nuclear reactor, some vibration tests are usually necessary to verify the prototype dynamical response characteristics and the structural integrity. To perform these tests, the known hydrodynamic loop facilities are used to evaluate the vibrational response of the bundle under the different flow conditions that may appear in the reactor. This paper describes the tests performed on a 19 plate fuel bundle prototype designed for a low power research reactor. The tests were done in order to know the dynamical characteristics of the plates and also of the whole bundle under different flow rate conditions. The paper includes a description of the test facilities and the results obtained during the dynamical characterization tests and some preliminary comments about the tests under flowing water are also presented. (author)

  8. Post-test examination of the VVER-1000 fuel rod bundle CORA-W2

    Hofmann, P.; Noack, V.; Burbach, J.; Metzger, H.; Schanz, G.; Hagen, S.; Sepold, L.

    1995-08-01

    The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B{sub 4}C absorber rod and the stainless steel grid spacers on the `low-temperature` bundle damage initiation and progression. The B{sub 4}C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occured. (orig./HP)

  9. A finite element model for static strength analysis of CANDU fuel bundle

    Horhoianu, G.; Ionescu, D.V. [Institute for Nuclear Research, Pitesti (Romania)

    2006-08-15

    A static strength analysis finite-element model has been developed using the ANSYS computer code in order to simulate the axial compression in CANDU type fuel bundle subject to hydraulic drag loads, deflection of fuel elements and stresses and displacements in the end plates. The validation of the finite-element model has been done by comparison with the out-reactor strength test results. Comparison of model predictions with the experimental results showed very good agreement. The comparative assessment reveals that SEU43 and SEU43L fuel bundles are able to withstand high flow rate without showing a significant geometric instability. (orig.)

  10. Total evaluation of in bundle void fraction measurement test of PWR fuel assembly

    Nuclear Power Engineering Corporation is performing the various proof or verification tests on safety and reliability of nuclear power plants under the sponsorship of the Ministry of International Trade and Industry. As one program of these Japanese national projects, an in bundle void fraction measurement test of a pressurized water reactor (PWR) fuel assembly was started in 1987 and finished at the end of 1994. The experiments were performed using the 5 x 5 square array rod bundle test sections. The rod bundle test section simulates the partial section and full length of a 17 x 17 type Japanese PWR fuel assembly. A distribution of subchannel averaged void fraction in a rod bundle test section was measured by the gamma-ray attenuation method using the stationary multi beam systems. The additional single channel test was performed to obtain the required information for the calibration of the rod bundle test data and the assessment of the void prediction method. Three test rod bundles were prepared to analyze an axial power distribution effect, an unheated rod effect, and a grid spacer effect. And, the obtained data were used for the assessment of the void prediction method relevant to the subchannel averaged void fraction of PWR fuel assemblies. This paper describes the outline of the experiments, the evaluation of the experimental data and the assessment of void prediction method

  11. Temperature Distributions in LMR Fuel Pin Bundles as Modeled by COBRA-IV-I

    Wright, Steven A.; Stout, Sherry

    2005-02-01

    Most pin type reactor designs for space power or terrestrial applications group the fuel pins into a number of relatively large fuel pin bundles or subassemblies. Fuel bundles for terrestrial liquid metal fast breeders reactors typically use 217 - 271 pins per sub-assembly, while some SP100 designs use up to 331 pins in a central subassembly that was surrounded by partial assemblies. Because thermal creep is exponentially related to temperature, small changes in fuel pin cladding temperature can make large differences in the lifetime in a high temperature liquid metal reactor (LMR). This paper uses the COBRA-IV-I computer code to determine the temperature distribution within LMR fuel bundles. COBRA-IV-I uses the sub-channel analysis approach to determine the enthalpy (or temperature) and flow distribution in rod bundles for both steady-state and transient conditions. The COBRA code runs in only a few seconds and has been benchmarked and tested extensively over a wide range of flow conditions. In this report the flow and temperature distributions for two types of lithium cooled space reactor core designs were calculated. One design uses a very tight fuel pin packing that has a pitch to diameter ratio of 1.05 (small wire wrap with a diameter of 392 μm) as proposed in SP100. The other design uses a larger pitch to diameter ratio of 1.09 with a larger more conventional sized wire wrap diameter of 1 mm. The results of the COBRA pin bundle calculations show that the larger pitch-to-diameter fuel bundle designs are more tolerant to local flow blockages, and in addition they are less sensitive to mal-flow distributions that occur near the edges of the subassembly.

  12. Flow-induced vibration and acoustic behaviour of CANFLEX-LVRF bundles in a Bruce B NGS fuel channel

    Frequency/temperature sweep tests were performed in a high-temperature/high-pressure test channel to determine the acoustic and flow-induced vibration characteristics of the CANFLEX-LVRF bundle. The vibratory response of CANFLEX-LVRF bundles was compared with that of 37-element fuel bundles under Bruce B NGS fuel channel normal operating conditions. The tests were performed with a 12-bundle string of CANFLEX-LVRF bundles as well as a mixed string for the transition core. The tests showed that the LVRF bundles performed as required without failure or gross geometry changes. The mixed fuel strings behaved in a manner similar to that of a string of CANFLEX-LVRF bundles. (author)

  13. Measurement and CFD calculation of spacer loss coefficient for a tight-lattice fuel bundle

    Highlights: • Experiment and CFD analysis evaluated the pressure drop in a spacer grid. • The measurement and CFD errors for the spacer loss coefficient were estimated. • The spacer loss coefficient for the dual-cooled annular fuel bundle was determined. • The CFD prediction agrees with the measured spacer loss coefficient within 8%. - Abstract: An experiment and computational fluid dynamics (CFD) analysis were performed to evaluate the pressure drop in a spacer grid for a dual-cooled annular fuel (DCAF) bundle. The DCAF bundle for the Korean optimum power reactor (OPR1000) is a 12 × 12 tight-lattice rod array with a pitch-to-diameter ratio of 1.08 owing to a larger outer diameter of the annular fuel rod. An experiment was conducted to measure the pressure drop in spacer grid for the DCAF bundle. The test bundle is a full-size 12 × 12 rod bundle with 11 spacer grid. The test condition covers a Reynolds number range of 2 × 104–2 × 105 by changing the temperature and flow rate of water. A CFD analysis was also performed to predict the pressure drop through a spacer grid using the full-size and partial bundle models. The pressure drop and loss coefficient of a spacer grid were predicted and compared with the experimental results. The CFD predictions of spacer pressure drop and loss coefficient agree with the measured values within 8%. The spacer loss coefficient for the DCAF bundle is estimated to be approximately 1.50 at a nominal operating condition of OPR1000, i.e., Re = 4 × 105

  14. Visual observations of a degraded bundle of irradiated fuel: the Phebus FPT1 test

    The international Phebus-FP (Fission Product) project is managed by the Institut de Protection et Surete Nucleaire in collaboration with Electricite de France (EDF), the European Commission (EC), the USNRC (USA), COG (Canada), NUPEC and JAERI (Japan), KAERI (South Korea), PSI and HSK (Switzerland). It is designed to measure the source-term and to study the degradation of irradiated UO2 fuel in conditions typical of a severe loss of coolant accident in a pressurised water reactor (PWR). In the first test (FPT0), performed in December '93, a bundle of 20 fresh fuel rods and a central Ag-In-Cd control rod underwent a short 15-day irradiation to generate fission products before testing in the Phebus reactor in Cadarache. The second test (FPT1) was performed in July '96, in the same conditions and geometry, but using irradiated fuel (-23 GWd/tU). In the FPT1 test, the bundle was heated to an estimated 3000 K over a period of 30 minutes in order to induce a substantial liquefaction of the bundle. After the test, the bundle was embedded in epoxy and cut at different levels to investigate the mechanisms of the core degradation. This paper reports the visual observations of the degraded FPT1 bundle, very preliminary interpretations about the scenario of degradation and a comparison between the behaviour of the fuel in the FPT0 and FPT1 tests. (author)

  15. Post-irradiation examination of CANDU fuel bundles fuelled with (Th, Pu)O2

    AECL has extensive experience with thoria-based fuel irradiations as part of an ongoing R&D program on thorium within the Advanced Fuel Cycles Program. The BDL-422 experiment was one component of the thorium program that involved the fabrication and irradiation testing of six Bruce-type bundles fuelled with (Th, Pu)O2 pellets. The fuel was manufactured in the Recycle Fuel Fabrication Laboratories (RFFL) at Chalk River allowing AECL to gain valuable experience in fabrication and handling of thoria fuel. The fuel pellets contained 86.05 wt. % Th and 1.53 wt. % Pu in (Th, Pu)O2. The objectives of the BDL-422 experiment were to demonstrate the ability of 37-element geometry (Th, Pu)O2 fuel bundles to operate to high burnups up to 1000 MWh/kgHE (42 MWd/kgHE), and to examine the (Th, Pu)O2 fuel performance. This paper describes the post-irradiation examination (PIE) results of BDL-422 fuel bundles irradiated to burnups up to 856 MWh/kgHE (36 MWd/kgHE), with power ratings ranging from 52 to 67 kW/m. PIE results for the high burnup bundles (>1000 MWh/kgHE) are being analyzed and will be reported at a later date. The (Th, Pu)O2 fuel performance characteristics were superior to UO2 fuel irradiated under similar conditions. Minimal grain growth was observed and was accompanied by benign fission gas release and sheath strain. Other fuel performance parameters, such as sheath oxidation and hydrogen distribution, are also discussed. (author)

  16. Post-irradiation examination of CANDU MOX fuel bundle containing weapons grade plutonium

    Dimayuga, F.C.; Karam, M.; Montin, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2008-07-01

    The Parallex Project is an experiment designed to demonstrate the feasibility of dispositioning US and Russian weapons grade plutonium (WPu) in CANDU reactors as a mixed-oxide (MOX) fuel. The Parallex Project involved the fabrication, irradiation testing, and post-irradiation examination (PIE) of three experimental CANDU MOX fuel bundles containing WPu fuel elements that were manufactured in the US and Russia. Some of the bundles contained MOX fuel fabricated at Chalk River Laboratories (CRL) from civilian plutonium (CivPu). This paper will describe the irradiation testing and post-irradiation examination of the second Parallex bundle. The second Parallex bundle is a 37-element bundle with its centre element removed to accommodate its irradiation in the National Research Universal (NRU) reactor. The bundle was assembled at CRL using intermediate and inner elements containing WPu MOX fuel pellets fabricated by the Bochvar Institute (Russia) and CivPu MOX pellets fabricated by AECL. The 18 outer elements were fuelled with natural uranium oxide fuel pellets containing dysprosia (to reduce the neutron flux that the Pu-bearing elements would be exposed to). Half of the intermediate and inner elements contained MOX fuel pellets fabricated with depleted uranium containing 4.6 wt% WPu. The other half of the intermediate and inner elements contained MOX fuel pellets fabricated with depleted uranium containing 5.3 wt% CivPu. The irradiation testing of the second bundle was completed in NRU. The intermediate MOX elements experienced linear powers up to 49 kW/m and achieved a burnup of 294 MWh/kgHE (12 MWd/kgHE). The inner MOX elements experienced linear powers up to 23 kW/m and achieved a burnup of 130 MWh/kgHE (5 Wd/kgHE). There was a significant difference between the performance of AECL-made MOX fuel containing CivPu and Russian MOX fuel containing WPu in terms of fission gas release (FGR). This is attributed to the different fabrication processes used to manufacture the

  17. Input modelling of ASSERT-PV V2R8M1 for RUFIC fuel bundle

    This report describes the input modelling for subchannel analysis of CANFLEX-RU (RUFIC) fuel bundle which has been developed for an advanced fuel bundle of CANDU-6 reactor, using ASSERT-PV V2R8M1 code. Execution file of ASSERT-PV V2R8M1 code was recently transferred from AECL under JRDC agreement between KAERI and AECL. SSERT-PV V2R8M1 which is quite different from COBRA-IV-i code has been developed for thermalhydraulic analysis of CANDU-6 fuel channel by subchannel analysis method and updated so that 43-element CANDU fuel geometry can be applied. Hence, ASSERT code can be applied to the subchannel analysis of RUFIC fuel bundle. The present report was prepared for ASSERT input modelling of RUFIC fuel bundle. Since the ASSERT results highly depend on user's input modelling, the calculation results may be quite different among the user's input models. The objective of the present report is the preparation of detail description of the background information for input data and gives credibility of the calculation results

  18. IFPE/AECL-BUNDLE, Fission Gas Release and Burnup Analysis, PHWR Fuel

    Description: Prototype Candu Fuel bundles for the CANDU6 (bundle NR) and Bruce (bundle JC) reactors were irradiated in the NRU experimental reactor at Chalk River Laboratories in experimental loop facilities under typical Candu reactor conditions, except that they were cooled using light water. NEA-1596/01 - Description: Bundle JC was a prototype 37-element fuel bundle for the Bruce-A Ontario Hydro reactors. This pressurized heavy water reactor (PHWR) design utilizes a heavy water moderator and pressurize heavy water coolant. For irradiation in the NRU reactor, the centre fuel element was removed and replaced by a central tie rod for irradiation purposes in the vertical test section. Coolant for the test was pressurized light water under typical PHWR conditions of 9 to 10.5 MPa and 300 deg. C. The fuel elements used 1.55 wt% U-235 in U uranium dioxide fuel and were clad with Zircaloy-4 material. The bundles' elements were coated with a graphite coating. The fuel is somewhat atypical of 37 element-type fuel since the length to diameter ratio (l/d) is large (1.73) due to the pellets being ground down from a OD of 14.3 mm to 12.12 mm. The outer element burnup averaged approximately 640 MWh/kgU on discharge. Outer element powers varied between 57 kW/m near the beginning of life and 23 kW/m at discharge. Due to the long irradiation, the bundle experienced 153 short shutdowns, and 129 longer duration shutdowns. No element instrumentation was used during the irradiation. However, the bundle was subjected to extensive post-irradiation examination (PIE) that included dimensional changes, fission gas release, fuel burnup analysis, and metallography that included grain size measurement. NEA-1596/02 - Description: Bundle NR was a prototype 37-element fuel bundle for the Candu 600 reactor. This pressurized heavy water reactor (PHWR) design utilizes a heavy water moderator and pressurized heavy water coolant. For irradiation in the NRU reactor, the centre fuel element was

  19. Metallographic examination of a CANDU fuel bundle heated under severe accident conditions

    Post-test metallographic examination of bundle cross sections of a 19-element modified CANDU fuel bundle was carried out. The bundle, HTBS-004, had been subjected to a severe temperature excursion to 1900 degrees Celsius in superheated steam. For this study, quantitative image analysis, Auger analysis and SEM-EDX techniques were applied. A significantly large quantity of molten (Zr, U, O) alloy was relocated in the bundle section 50 mm from the upstream end, whereas the 377-mm section showed little relocated material except at the inner element junctions. These variations in the molten material generation and relocation have been correlated with the corresponding axial and radial variations in the heatup rates

  20. Posttest examination of the VVER-1000 fuel rod bundle CORA-W2

    The bundle meltdown experiment CORA-W2, representing the behavior of a Russian type VVER-1000 fuel element, with one B4C/stainless steel absorber rod was selected by the OECD/CSNI as International Standard Problem (ISP-36). The experimental results of CORA-W2 serve as data base for comparison with analytical predictions of the high-temperature material behavior by various code systems. The first part of the experimental results is described in KfK 5363 (1994), the second part is documented in this report which contains the destructive post-test examination results. The metallographical and analytical (SEM/EDX) post-test examinations were performed in Germany and Russia and are summarized in five individual contributions. The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B4C absorber rod and the stainless steel grid spacers on the ''low-temperature'' bundle damage initiation and progression. The B4C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occurred. (orig.)

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

    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)

  2. Results of international standard problem No. 36 severe fuel damage experiment of a VVER fuel bundle

    International Standard Problems (ISP) organized by the OECD are defined as comparative exercises in which predictions with different computer codes for a given physical problem are compared with each other and with a carefully controlled experimental study. The main goal of ISP is to increase confidence in the validity and accuracy of analytical tools used in assessing the safety of nuclear installations. In addition, it enables the code user to gain experience and to improve his competence. This paper presents the results and assessment of ISP No. 36, which deals with the early core degradation phase during an unmitigated severe LWR accident in a Russian type VVER. Representatives of 17 organizations participated in the ISP using the codes ATHLET-CD, ICARE2, KESS-III, MELCOR, SCDAP/RELAP5 and RAPTA. Some participants performed several calculations with different codes. As experimental basis the severe fuel damage experiment CORA-W2 was selected. The main phenomena investigated are thermal behavior of fuel rods, onset of temperature escalation, material behavior and hydrogen generation. In general, the calculations give the right tendency of the experimental results for the thermal behavior, the hydrogen generation and, partly, for the material behavior. However, some calculations deviate in important quantities - e.g. some material behavior data - showing remarkable discrepancies between each other and from the experiments. The temperature history of the bundle up to the beginning of significant oxidation was calculated quite well. Deviations seem to be related to the overall heat balance. Since the material behavior of the bundle is to a great extent influenced by the cladding failure criteria a more realistic cladding failure model should be developed at least for the detailed, mechanistic codes. Regarding the material behavior and flow blockage some models for the material interaction as well as for relocation and refreezing requires further improvement

  3. Results of international standard problem No. 36 severe fuel damage experiment of a VVER fuel bundle

    Firnhaber, M. [Gesellschaft fuer Anlagen-und Reaktorsicherheit, Koeln (Germany); Yegorova, L. [Nuclear Safety Institute of Russian Research Center, Moscow (Russian Federation); Brockmeier, U. [Ruhr-Univ. of Bochum (Germany)] [and others

    1995-09-01

    International Standard Problems (ISP) organized by the OECD are defined as comparative exercises in which predictions with different computer codes for a given physical problem are compared with each other and with a carefully controlled experimental study. The main goal of ISP is to increase confidence in the validity and accuracy of analytical tools used in assessing the safety of nuclear installations. In addition, it enables the code user to gain experience and to improve his competence. This paper presents the results and assessment of ISP No. 36, which deals with the early core degradation phase during an unmitigated severe LWR accident in a Russian type VVER. Representatives of 17 organizations participated in the ISP using the codes ATHLET-CD, ICARE2, KESS-III, MELCOR, SCDAP/RELAP5 and RAPTA. Some participants performed several calculations with different codes. As experimental basis the severe fuel damage experiment CORA-W2 was selected. The main phenomena investigated are thermal behavior of fuel rods, onset of temperature escalation, material behavior and hydrogen generation. In general, the calculations give the right tendency of the experimental results for the thermal behavior, the hydrogen generation and, partly, for the material behavior. However, some calculations deviate in important quantities - e.g. some material behavior data - showing remarkable discrepancies between each other and from the experiments. The temperature history of the bundle up to the beginning of significant oxidation was calculated quite well. Deviations seem to be related to the overall heat balance. Since the material behavior of the bundle is to a great extent influenced by the cladding failure criteria a more realistic cladding failure model should be developed at least for the detailed, mechanistic codes. Regarding the material behavior and flow blockage some models for the material interaction as well as for relocation and refreezing requires further improvement.

  4. Measurement of Quasi-periodic Oscillating Flow Motion in Simulated Dual-cooled Annular Fuel Bundle

    In order to increase a significant amount of reactor power in OPR1000, KAERI (Korea Atomic Energy Research Institute) has been developing a dual-cooled annular fuel. The dual-cooled annular fuel is simultaneously cooled by the water flow through the inner and the outer channels. KAERI proposed the 12x12 dual-cooled annular fuel array which was designed to be structurally compatible with the 16x16 cylindrical solid fuel array by maintaining the same array size and the guide tubes in the same locations, as shown in Fig. 1. In such a case, due to larger outer diameter of dual-cooled annular fuel than conventional solid fuel, a P/D (Pitch-to-Diameter ratio) of dual cooled annular fuel assembly becomes smaller than that of cylindrical solid fuel. A change in P/D of fuel bundle can cause a difference in the flow mixing phenomena between the dual-cooled annular and conventional cylindrical solid fuel assemblies. In this study, the rod bundle flow motion appearing in a small P/D case is investigated preliminarily using PIV (Particle Image Velocimetry) for dual-cooled annular fuel application

  5. Prediction of temperature distribution in a fast reactor spent fuel bundle

    A simple mathematical model is described for predicting temperature distribution in a spent fuel bundle. The model takes into account γ-ray leakage, radiant and conductive heat transports between the various fuel pins arranged in a triangular array and enclosed in a hexagonal shaped tube containing gaseous medium. With the geometry of the fuel bundle the configuration factors between various fuel pins can be calculated. The configuration factors along with the heat generation rates, net γ-ray leakage, surface emissivity, conductivity of the enclosed medium and the temperature of the hexagonal tube can be used to estimate the temperature distribution with the help of the computer code TICOFUSA developed on the basis of this model. (author)

  6. An analytical method for predicting the temperature distribution in an irradiated fuel pin bundle

    A simple analytical model is described for predicting the temperature distribution in a spent fuel bundle. The model takes into account gamma-ray transport, radiant and conductive heat transports between the various fuel pins arranged in a triangular array and enclosed in a hexagonal shaped tubes containing gaseous medium. With the geometry of the fuel bundle the configuration factors between various fuel pins can be calculated from the relations presented in this report. The configuration factors along with the heat generation rates, net gamma ray leakage, surface emissivity, conductivity of the enclosed medium and the temperature of the hexagonal tube can be used to estimate the temperature distribution with the help of the computer code developed on the basis of this model. (orig.)

  7. Status of the demonstration irradiation of the CANDU new fuel bundle CANFLEX-NU in Korea

    A demonstration irradiation (DI) of 24 KNFC made CANFLEX-NU fuel bundles in the Wolsong Power Generation Station-i has been conducted jointly by KEPRI/KHNP/KAERI since July 10, 2002. By selecting the Q07 (high power) and L21(low power) channels, the total 24 and 16 CANFLEX bundles were respectively loaded into and discharged from the reactor by 2003 August, and the final discharge of the other 8 CANFLEX bundles is expected on around February 2004. Tracking the reactor operation data, it is noted that the reactor has been stably operated during the DI. One CANFLEX bundle irradiated in the Q07 channel had a typical history of high power and high burnup, having the outer element power rating of ∼ 41 kW/m at the fuelling, ∼ 42 kW/m as a maximum power rating at the burnup of ∼ 50 MWh/kgU, and ∼ 35 kW/m at the discharge burnup of ∼ 210 MWh/kgU. While, another CANFLEX bundle also irradiated in the Q07 channel had a typical history of power ramping, having a outer element power rating of ∼ 7 kW/m from the fuelling to the burnup of ∼ 48 MWh/kgU at which the element powers were ramped to a ∼ 35 kW/m maximum element power rating, and ∼ 30 kW/m at the discharge burnup of 188 MWh/kgU. An unusual performance and integrity of the CANFLEX elements could not be found in the ELESTRES predictions. By looking at the discharged CANFLEX bundles in the bay, all the bundles were intact, free of defects and appeared to be in good condition. A detailed in-bay visual examinations and dimensional measurements of the discharged CANFLEX bundles will be made at the end of 2003. (author)

  8. Candu reactors with thorium fuel cycles

    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 5500C. 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

  9. Calculation of power coefficient in CANFLEX-NU fuel bundle

    Min, Byung Joo; Jun, Ji Su; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-11-01

    Changes in power level affect reactivity due to its dependence on fuel and coolant temperatures. The power coefficient of reactivity is related to the fuel temperature coefficient through the change in fuel temperature per percent change in power. In addition, power level changes are followed by changes in coolant temperature and density which contribute to the reactivity effect. In this report, the power coefficient of CANFLEX-NU was calculated and the result would be compared with that of CANDU-6 reactor which is operating. 8 refs., 43 figs., 2 tabs. (Author)

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

    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

  11. Optimising welding and assembling processes for manufacturing PHWR fuel element and bundle

    In PHWR fuel fabrication, end-cap joint formed by Zircaloy fuel tube and cap is one of the most critical welds as it is expected to offer a hermetically sealed joint to contain the radioactive fission products. In view of their highly demanding function during reactor operation, these welds have to be produced to a high degree of reliability by careful selection of process and parameters. PHWR fuel bundle is manufactured by joining end plates to elements at both ends. Resistance projection welding technique is used to weld the element ends to end plates. This being the final operation in PHWR fuel fabrication route, it plays very important role with respect to bundle dimensions and integrity. Jigs and Fixtures are used to assemble fuel elements and end plates. The quality of these fixtures affect the bundle dimensions, inter element spacing and orientation of fuel elements/end-plates. While welding Zircaloy material, properties like coefficient of thermal expansion, thermal conductivity and thin oxide layers have to be considered. Generally high conductive material requires pre-heating before welding, while post-treatment of the weld is carried out if the metallurgical properties are changing in the Heat Affected Zone (HAZ). In resistance welding, selecting a suitable weld cycle pattern involves optimization of current, time, number of on/off cycles and current slope. Different current cycle patterns offer distinct advantages and certain disadvantages too with respect to weld bonding, sparking, HAZ etc. State-of-the-art technology is being used to have better control on weld parameters and monitor them as well for further analysis. The paper discusses the effect of welding parameters including different weld cycle patterns like on/off cycle, up-slope cycle and constant current cycle. Improvements carried out to ensure dimensional integrity of the bundle are also dealt with in the paper. (author)

  12. A study of coolant thermal mixing within CANDU fuel bundles using ASSERT-PV

    This paper presents the results of a study of the thermal mixing of single-phase coolant in 28-element CANDU fuel bundles. The approach taken in the present work is to identify the physical mechanisms contributing to coolant mixing, and to systematically assess the importance of each mechanism. Coupled effects were also considered by flow simulation with mixing mechanisms modelled simultaneously. For the limited range of operating conditions considered and when all mixing mechanisms were modelled simultaneously, the flow was found to be very close to fully mixed. A preliminary model of coolant mixing, suitable for use in the fuel and fuel channel code FACTAR, is also presented. (author)

  13. Exceptional crud build-up in Loviisa-2 fuel bundles

    Anomalous primary coolant outlet temperatures at Loviisa 2 unit were first discovered in October, 1994, one month after the start of the 15. cycle. The reason for increased outlet temperatures was soon found out to be decreased coolant flow through part of the fuel assemblies. This phenomenon was most pronounced in six first cycle fuel assemblies with spacer grids made of Zr1%Nb (ZR assemblies). Due to continuously increasing outlet temperature the reactor was shut down at the end of January, 1995. The six ZR assemblies were discharged from the reactor. Towards the end of cycle no. 15 the rate of outlet temperature increase slowed down and essentially stopped in the remaining assemblies, which had spacer grids made of stainless steel (SS assemblies). One of the ZR assemblies was visually inspected using the pool-side inspection equipment at Loviisa 2 unit. This inspection showed that the reason for the decreased coolant flow was deposition of crud in the spacer grids, especially in the lower parts of the assembly. Based on data of coolant outlet temperatures, flow resistance measurements were carried out for eighty SS assemblies during the refuelling outage between cycles no. 15 and no. 16. As a result thirty assemblies, which had the most clogged spacer grids, were discharged from the reactor before their planned end of life. The cycle no. 16 started with an indication of a small leakage in September, 1995. Primary coolant activity kept increasing steadily, indicating more fuel failures, up to values never reached before at Loviisa NPP. The estimated number of leaking rods varied from approximately 10 rods up to ca. 70 rods. Finally, Loviisa 2 unit was decided to be shut down in late October, 1995. Sipping of the core indicated that there were seven leaking fuel assemblies in the reactor. All leaking assemblies had earlier been identified as being slightly clogged due to the deposition of crud in the spacer grids. Altogether thirty-two slightly clogged assemblies

  14. The demonstration irradiation of the CANFLEX-NU fuel bundle in Wolsong NGS 1

    A demonstration irradiation (DI) of 24 CANFLEX-NU fuel bundles in the high power Q07 channel and low power L21 channel of Wolsong Power Generation Station-1 had been successfully conducted jointly by KEPRI/KHNP/KAERI in the period of 2002 July to 2004 January. The tracking of the reactor operation data showed that the reactor has been stably operated during the DI. One CANFLEX bundle irradiated in the Q07 channel had a typical history of high power and high burnup, where the maximum element linear power rating was ∼ 42 kW/m at the burnup of ∼ 50 MWh/kgU and ∼ 35 kW/m at the discharge element burnup of ∼ 210 MWh/kgU. While, another CANFLEX bundles also irradiated in the Q07 channel had a typical history of power ramping, where the maximum element power ramping-up or -down rate was 28 kW/m. The unusual performance and integrity of the CANFLEX elements could not be found in the ELESTRES predictions and also the in-bay visual examinations showed that all the bundles were intact, free of defects and appeared to be in good condition as expected. Therefore, it is concluded that the demonstration irradiation shows the validation of the CANFLEX bundle performance with direct conditions of relevance under the Korean licensing requirements and the KNFC fuel fabrication capability, and provides the rationale for the decision to perform the full-conversion of CANFLEX fuel in WPGS-1. (author)

  15. Numerical simulation of fluid flow and heat transfer of supercritical fluids in fuel bundles

    A supercritical water-cooled reactor (SCWR) was proposed as a kind of generation IV reactor in order to improve the efficiency of nuclear reactors. Although investigations on the thermal-hydraulic behavior in SCWR have attracted much attention, there is still a lack of CFD study on the heat transfer of supercritical water in fuel channels. In order to understand the thermal-hydraulic behavior of supercritical fluids in nuclear reactors, the local fluid flow and heat transfer of supercritical water in a 37-element fuel bundle has been studied numerically in this work. Results show that secondary flow appears and the cladding surface temperature (CST) is very nonuniform in the fuel bundle. The maximum cladding surface temperature (MaxCST), which is an important design parameter for SCWR, can be predicted and analyzed using the CFD method. Due to a very large circumferential temperature gradient in cladding surfaces of the fuel bundle, the precise cladding temperature distributions using the CFD method is highly recommended. (author)

  16. Development of neural network for analysis of local power distributions in BWR fuel bundles

    A neural network model has been developed to learn the local power distributions in a BWR fuel bundle. A two layers neural network with total 128 elements is used for this model. The neural network learns 33 cases of local power peaking factors of fuel rods with given enrichment distribution as the teacher signals, which were calculated by a fuel bundle nuclear analysis code based on precise physical models. This neural network model studied well the teacher signals within 1 % error. It is also able to calculate the local power distributions within several % error for the different enrichment distributions from the teacher signals when the average enrichment is close to 2 %. This neural network is simple and the computing speed of this model is 300 times faster than that of the precise nuclear analysis code. This model was applied to survey the enrichment distribution to meet a target local power distribution in a fuel bundle, and the enrichment distribution with flat power shape are obtained within short computing time. (author)

  17. Air-water two-phase flow pressure drop across various components of AHWR fuel bundle

    Single-phase (water) and two-phase (air-water) experiments were carried out for the measurement of pressure drops across various components of a prototype full scale 54-rod fuel bundle of proposed AHWR (Advanced Heavy Water Reactor). From the measured values of pressure drops, the friction factor for fuel bundle and the loss coefficients for the tie plates and spacers were estimated. The single-phase experimental data were compared with different existing correlations. Correlations have been proposed based on the data generated with the air-water mixture which can be used for prediction of pressure drop across fuel channel (with 54 rod fuel bundle) of AHWR under normal operating conditions with appropriate correction factor for steam-water flow. Also a heuristic approach to predict the Lockhart-Martinelli parameter has been presented. Further, a new correlation for two-phase friction multiplier applicable to 54-rod cluster geometry has been developed based on two-phase experimental pressure drop data. The effect of mixture mass flux on the two-phase friction multiplier has been probed and the assessment of existing friction multiplier correlations has also been carried out with the test data. (author)

  18. Status of the demonstration irradiation program of the new fuel bundle CANFLEX-NU in Korea

    In the late part of 1999, the Korea Electric Power Corporation has initiated a program CANFLEX-NU (Natural Uranium) fuel in the Wolsong Generating Station (WGS) - no.1 which has been operating since 1983, because the CANFLEX could be used to recover some of a CANDU heat transport system operation margins that had decreased due to The Korea Ministry of Science and Technology (MOST) has recognized the successful demonstration irradiation of 24 CANFLEX bundles at the Pt. Lepreau Generating Station in Canada, as final verification of the CANFLEX design in preparation for full core conversion. Therefore, MOST has pushed and gave a financial support to a KEPRI/KAERI Joint Industrialization Program of CANFLEX-NU Fuel, which will be for 3 years from 2000 November, to validate CANFLEX-NU fuel bundle performance in direct conditions of relevance under the Korean licensing requirements as well as to evaluate the fuel fabrication capability, and to produce a safety analysis report for the full-core implementation. The economic benefits of CANFLEX-NU fuel are directly dependent on the thermalhydraulic performance. Switching from the existing 37-element fuel to the CANFLEX fuel will be largely driven by the economic benefits to be realized. Showing a positive result in the economic evaluation as well as successfully demonstrating the CANFLEX fuel irradiation in WGS-no. 1, the full-core implementation of the fuel at the WGS-no.1 in Korea will proceed by starting the licensing process at around 2003 April because the safety report for the full-core conversion will be ready by 2003 March. This paper describes the status of CANFLEX-NU fuel industrialization program in Korea, as well as the fuel design features. It summarizes the plan of CANFLEX-NU fuel demonstration irradiation at the WGS-no. 1 in Korea and the status of documentation for the demonstration irradiation as well as for the CANFLEX-NU full-core implementation. (author)

  19. Development of neural network simulating power distribution of a BWR fuel bundle

    A neural network model is developed to simulate the precise nuclear physics analysis program code for quick scoping survey calculations. The relation between enrichment and local power distribution of BWR fuel bundles was learned using two layers neural network (ENET). A new model is to introduce burnable neutron absorber (Gadolinia), added to several fuel rods to decrease initial reactivity of fresh bundle. The 2nd stages three layers neural network (GNET) is added on the 1st stage network ENET. GNET studies the local distribution difference caused by Gadolinia. Using this method, it becomes possible to survey of the gradients of sigmoid functions and back propagation constants with reasonable time. Using 99 learning patterns of zero burnup, good error convergence curve is obtained after many trials. This neural network model is able to simulate no learned cases fairly as well as the learned cases. Computer time of this neural network model is about 100 times faster than a precise analysis model. (author)

  20. Analytical and CFD investigation of ex-core cooling of the nuclear fuel rod bundle in a water pool

    The efficiency of ex-core cooling of nuclear fuel assemblies under decay heat generation is influenced by many conditions, among them being coolant flow rate, position of fuel assemblies in a water pool, and position of coolant inlets and outlets. A combination of unacceptable thermal-hydraulic conditions occurred at the Nuclear Power Plant PAKS in Hungary in April 2003, during the process of nuclear fuel assembly chemical cleaning in a specially designed tank. The cooling of the nuclear fuel rod bundles in the tank was not efficient under low coolant flow rates through the cleaning tank, and after several hours the boiling of cooling water occurred with subsequent dry-out of nuclear fuel rod bundles. The thermal-hydraulic conditions in the cleaning tank that led to the unexpected event are analysed both analytically and with a CFD approach for idealized conditions of one nuclear fuel rod bundle with the bottom by-pass opening. The analytical analysis is based on a pressure balance of low Reynolds number upward water coolant flow through the bundle, downward water flow in the pool around the bundle, flow across the by-pass opening and outlet flow from the cleaning vessel. The transient CFD simulations are performed in order to demonstrate multidimensional effects of the event. The water density dependence on the temperature is taken into account in both analytical and CFD investigation, as the dominant effect that influences the buoyancy forces between the water flow streams inside and outside the vertically positioned bundle in the water pool. The influence of the bundle bottom by-pass area on the water pool thermal-hydraulic conditions and on the efficiency of the nuclear fuel rods cooling is analysed. Both analytical and CFD results show that the continuous cooling of the fuel rods can not be achieved for higher values of the bundle bottom by-pass areas. The averaged coolant temperature in the water pool outside the bundle becomes higher than the average

  1. Thermal-hydraulic stability tests for newly designed BWR rod bundle (step-III fuel type B)

    The Step-III Fuel Type B is a new fuel design for high burn-up operation in BWRs in Japan. The fuel design uses a 9x9 - 9 rod bundle to accommodate the high fuel duty of high burn-up operation and a square water-channel to provide enhanced neutron moderation. The objective of this study is to confirm the thermal-hydraulic stability performance of the new fuel design by tests which simulate the parallel channel configuration of the BWR core. The stability testing was performed at the NFI test loop. The test bundle geometry used for the stability test is a 3x3 heater rod bundle which has about 1/8 of the cross section area of the full size 9x9 - 9 rod bundle. Full size heater rods were used to simulate the fuel rods. For parallel channel simulation, a bypass channel with a 6x6 - 8 heater rod bundle was connected in parallel with the 3x3 rod bundle test channel. The stability test results showed typical flow oscillation features which have been described as density wave oscillations. The stationary limit cycle oscillation extended flow amplitudes to several tens of a percent of the nominal value, during which periodic dry-out and re-wetting were observed. The test results were used for verification of a stability analysis code, which demonstrated that the stability performance of the new fuel design has been conservatively predicted. (author)

  2. An experimental investigation of the temperature behavior of a CANDU 37-element spent fuel bundle with air backfill

    As part of the thermal analysis of a CANDU spent fuel dry storage system, a series of experiment has been conducted using a thermal mock-up of a simulated CANDU spent fuel bundle in a dry storage basket. The experimental system was designed to obtain the maximum fuel rod temperature along with the radial and axial temperature distributions within the fuel bundle. The main purpose of these experiments was to characterize the relevant heat transfer mechanisms in a dry, vertically oriented CANDU spent fuel bundle, and to verify the MAXROT code developed for the thermal analysis of a CANDU spent fuel bundle in a dry storage basket. A total of 48 runs were made with 8 different power inputs to the 37-element heater rod bundle ranging from 5 to 40 W, while using 6 different band heaters power inputs from 0 to 250 W to maintain the basket wall at a desired boundary condition temperature at the steady state. The temperature distribution in a heater rod bundle was measured and recorded at the saturated condition for each set of heater rod power and band heaters power. To characterize the heat transfer mechanism involved, the experimental data were corrected analytically for radiation heat transfer and presented as a Nusselt number correlation in terms of the Rayleigh number of the heater rod bundle. The results show that the Nusselt number remains nearly constant and all the experimental dada fall within a conduction regime. The experimental data were compared with the predictions of the MAXROT code to examine the code's accuracy and validity of assumptions used in the code. The MAXROT code explicitly models each representative fuel rod in a CANDU fuel bundle and couples the conductive and radiative heat transfer of the internal gas between rods. Comparisons between the measured and predicted maximum fuel rod temperatures of the simulated CANDU 37-element spent fuel bundle for all 48 tests show that the MAXROT code slightly over-predicts and the agreement is within 2

  3. Computer code TOBUNRAD for PWR fuel bundle heat-up calculations

    The computer code TOBUNRAD developed is for analysis of ''fuel-bundle'' heat-up phenomena in a loss-of-coolant accident of PWR. The fuel bundle consists of fuel pins in square lattice; its behavior is different from that of individual pins during heat-up. The code is based on the existing TOODEE2 code which analyzes heat-up phenomena of single fuel pins, so that the basic models of heat conduction and transfer and coolant flow are the same as the TOODEE2's. In addition to the TOODEE2 features, unheated rods are modeled and radiation heat loss is considered between fuel pins, a fuel pin and other heat sinks. The TOBUNRAD code is developed by a new FORTRAN technique which makes it possible to interrupt a flow of program controls wherever desired, thereby attaching several subprograms to the main code. Users' manual for TOBUNRAD is presented: The basic program-structure by interruption method, physical and computational model in each sub-code, usage of the code and sample problems. (author)

  4. Utilization of fluorescent uranium x-rays as verification tool for irradiated CANDU fuel bundles

    The use of fluorescent uranium x-rays for in-situ safeguards verification of irradiated CANDU fuel bundles is described. Room temperature CdZnTe (supergrade) semiconductor detector of low sensitivity coupled to charge sensitive pre-amplifier is used. This detector is characterized by moderate resolving power in the low energy region around 100 keV. It as such allows the separation of uranium x-rays in the close proximity of tungsten x-rays emanating from the shielding/collimator assembly. On account of strong attenuation, the detection of low energy x-rays requires the shielding to be of an optimized thickness. Further, in view of high intensity of this radiation the use of small volume detector is warranted. In dealing with the subject, this paper therefore presents an assessment, not only of the detector but also the shield-collimator assembly for the required verification of short cooling time fuel bundles. Results of the associated optimization measurements with respect to collimator aperture and detector sensitivity are consequently included. The future course of work from the viewpoint of development of a suitable x-ray spectrometer specifically for the purpose of verifying extremely short (< 1 month old) cooling time fuel bundles is moreover identified. (author)

  5. The EC6 - an enhanced mid-sized reactor with fuel cycle applications

    Atomic Energy of Canada Limited (AECL) has two CANDU reactor products matched to markets: the Enhanced CANDU 6 (EC6), a modern 700 MWe-class design, and the Advanced CANDU Reactor (ACR-1000), a 1200 MWe-class Gen III+ design. Both reactor types are designed to meet both market-, and customer-driven needs; the ACR-1000 design is 90% complete and market-ready. The EC6 incorporates the CANDU 6's well-proven features, and adds enhancements that make the reactor even safer and easier to operate. The EC6 is the only mid-sized reactor with a proven pedigree that meets modern reactor expectations and regulatory standards. It is sized for smaller grids and also has outstanding fuel-cycle capability. The EC6 has domestic and offshore market pull and is the current focus of AECL's development program; market interest in the ACR-1000 is anticipated in the longer term. Some of the key features incorporated into the EC6 include upgrading containment and seismic capability to meet modern standards, shortening the overall project schedule, addressing obsolescence issues, optimizing maintenance outages and incorporating lessons learnt through feedback obtained from the operating plants. The EC6 utilizes modern computers and a distributed control system housed in an advanced control room which, along with automated testing and on-line diagnostics, make the plant easier and safer to operate, with minimal operator intervention. The first deployment of the EC6 is anticipated in Canada; off-shore markets are also being pursued. The EC6 burns natural uranium as standard. But, high neutron economy, on-power refuelling, a simple fuel bundle, and the fundamental CANDU fuel channel design provide the EC6 with the flexibility to accommodate a range of advanced fuels. (author)

  6. Resistance factors, two phase multipliers and void fractions for best estimate flow calculations in Dodewaard fuel bundles

    Values are given for resistance factors, two phase multipliers and core and chimney void fractions in the fuel and chimney to be used in best estimate calculations of the flow in Dodewaard fuel bundles. The resistance factors are based on single phase experimental data for a mockup of the Dodewaard fuel bundle. The two phase multipliers are determined from two phase measurements of mockups of other fuel bundles for nuclear reactors. This is also true for the in bundle void fractions. The void fractions in the chimney have been validated by measured void fractions in large diameter pipes. The recommended changes to the existing input for calculations are somewhat larger than the uncertainties in the measurements. (author). 37 refs.; 48 figs.; 4 tabs

  7. Advanced CFD simulations of turbulent flows around appendages in CANDU fuel bundles

    Computational Fluid Dynamics (CFD) was used to simulate the coolant flow in a modified 37-element CANDU fuel bundle, in order to investigate the effects of the appendages on the flow field. First, a subchannel model was created to qualitatively analyze the capabilities of different turbulence models such as k.ε, Reynolds Normalization Group (RNG), Shear Stress Transport (SST) and Large Eddy Simulation (LES). Then, the turbulence model with the acceptable quality was used to investigate the effects of positioning appendages, normally used in CANDU 37-element Critical Heat Flux (CHF) experiments, on the flow field. It was concluded that the RNG and SST models both show improvements over the k.ε method by predicting cross flow rates closer to those predicted by the LES model. Also the turbulence effects in the k.ε model dissipate quickly downstream of the appendages, while in the RNG and SST models appear at longer distances similar to the LES model. The RNG method simulation time was relatively feasible and as a result was chosen for the bundle model simulations. In the bundle model simulations it was shown that the tunnel spacers and leaf springs, used to position the bundles inside the pressure tubes in the experiments, have no measureable dominant effects on the flow field. The flow disturbances are localized and disappear at relatively short streamwise distances. (author)

  8. Use of radiography to monitor structural movement in GCFR-CFTL fuel rod bundles

    The Core Flow Test Loop (CFTL) is designed to simulate accident conditions of the Gas-Cooled Fast Reactor (GCFR). The reactor fuel rods are simulated by electric heater rods. An important consideration in data acquisition for loss of coolant studies is structural movement in the test bundle, that is, axial expansion and laterial movement (bowing) of fuel rod simulators and ducts. Radiography is superior to proximity sensors and extensometers for monitoring structural movement because radiography is external to the CFTL vessel and nonintrusive. Both fluoroscopy and film radiography were investigated. Both techniques were determined feasible, and both are recommended for GCFR-CFTL applications

  9. Out-of-pile bundle experiments on severe fuel damage (CORA-program): Objectives, test matrix and facility description

    As part of the Severe Fuel Damage Program by the German Nuclear Safety Project, out-of-pile experiments are being conducted at the Kernforschungszentrum Karlsruhe to investigate the damage behaviour of PWR fuel rod bundles under Severe Fuel Damage conditions (CORA-Program). This report describes the objectives, the test matrix and the CORA-facility. (orig.)

  10. Study of thermal hydraulic behavior of supercritical water flowing through fuel rod bundles

    Investigations on thermal-hydraulic behavior in Supercritical Water Reactor (SCWR) fuel assembly have obtained a significant attention in the international SCWR community because of its potential to obtain high thermal efficiency and compact design. Present work deals with CFD analysis to study the flow and heat transfer behavior of supercritical water in 4 metre long 7-pin fuel bundle using commercial CFD package ANSYS CFX for single phase steady state conditions. Considering the symmetric conditions, 1/12th part of the fuel rod bundle is taken as a domain of analysis. RNG K-epsilon model with scalable wall functions is used for modeling the turbulence behavior. Constant heat flux boundary condition is applied at the fuel rod surface. IAPWS equations of state are used to compute thermo-physical properties of supercritical water. Sharp variations in its thermo-physical properties (specific heat, density) are observed near the pseudo-critical temperature causing sharp change in heat transfer coefficient. The pseudo-critical point initially appears in the gaps among heated fuel rods, and then spreads radially outward reaching the adiabatic wall as the flow goes downstream. The enthalpy gain in the centre of the channel is much higher than that in the wall region. Non-uniformity in the circumferential distribution of surface temperature and heat transfer coefficient is observed which is in agreement with published literature. Heat transfer coefficient is high on the rod surface near the tight region and decreases as the distance between rod surfaces increases. (author)

  11. Thermal-hydraulic stability tests for newly designed BWR rod bundle (step-III fuel type A)

    Thermal-hydraulic stability tests have been performed on electrically heated bundles to simulate the newly designed Boiling Water Reactor (BWR) fuels in a parallel channel test loop. The objective of the current experimental program is to investigate how the newly designed bundle could improve the thermal-hydraulic stability. Measurements of the thermal-hydraulic instability thresholds in two vertical rod bundles have been conducted in steam-water two-phase flow conditions at the TOSHIBA test loop. Fluid conditions were BWR operating conditions of 7 MPa system pressure, 1.0-2.0x106 kg/m2/h inlet mass flux and 28-108 kJ/kg inlet subcooling. The parallel channel test loop consists of a main bundle of 3x3 indirectly heated rods of 1/9 symmetry of 9x9 full lattice and a bypass bundle of 8x8. These are both simulated BWR rod bundles in respect of rod diameter, heated length, rod configuration, fuel rod spacer, core inlet hydraulic resistance and upper tie plate. There are three types of the 3x3 test bundles with different configurations of a part length rod of two-thirds the length of the other rods and an axial power shape. The design innovation of the part length rod for a 9x9 lattice development, though addition of more fuel rods increases bundle pressure drop, reduces pressure drop in the two-phase portion of the bundle, and enhances the thermal hydraulic stability. Through the experiments, the parameter dependency on the channel stability threshold is obtained for inlet subcooling, inlet mass flux, inlet flow resistance, axial power shape and part length rod. The main conclusion is that the stability threshold is about 10% greater with the part length rod than without the part length rod. The new BWR bundle consisting of the part length rod has been verified in respect of thermal hydraulic stability performance. (author)

  12. Optimization of thorium-uranium content in a 54-element fuel bundle for use in a CANDU-SCWR

    A new 54-element fuel bundle design has been proposed for use in a pressure-tube supercritical water-cooled reactor, a pre-conceptual evolution of existing CANDU reactors. Pursuant to the goals of the Generation IV International Forum regarding advancement in nuclear fuel cycles, optimization of the thorium and uranium content in each ring of fuel elements has been studied with the objectives of maximizing the achievable fuel utilization (burnup) and total thorium content within the bundle, while simultaneously minimizing the linear element ratings and coolant void reactivity. The bundle was modeled within a reactor lattice cell using WIMS-AECL, and the uranium and thorium content in each ring of fuel elements was optimized using a weighted merit function of the aforementioned criteria and a metaheuristic search algorithm. (author)

  13. Description and validation of ANTEO, an optimised PC code the thermalhydraulic analysis of fuel bundles

    The paper deals with the description of a Personal Computer oriented subchannel code, devoted to the steady state thermal hydraulic analysis of nuclear reactor fuel bundles. The development of such a code was made possible by two facts: firstly, the increase, in the computing power of the desk machines; secondly, the fact that several years of experience into operate subchannels codes have shown how to simplify many of the physical models without a sensible loss of accuracy. For sake of validation, the developed code was compared with a traditional subchannel code, the COBRA one. The results of the comparison show a very good agreement between the two codes. (author)

  14. SAGAPO. A computer code for the thermo-fluiddynamic analysis of gas cooled fuel element bundles

    This paper is a guide for the users of the Fortran computer code SAGAPO, which has been developed by the author for the thermo-fluiddynamic analysis of gas cooled fuel element bundles. The physical models and the mathematical procedures used in SAGAPO have been already described by the author of this work in a previous paper. Thus this work contains only a description of the structure of the code, together with the other informations necessary to the users. A listing of SAGAPO is included in the appendix, together with an example of input preparation and parts of printed results. (orig.)

  15. Selection of instruments used for vibration measurement of fuel bundles in a pressure tube under CANDU reactor operating conditions

    Vibration characteristics of CANDU fuel bundle and fuel elements is a key parameter considered in the design of a fuel bundle. Out-reactor frequency and temperature sweep tests, under reactor operating conditions, are performed to verify vibration characteristics of CANDU fuel bundles. Several options have been considered in the selection of vibration instrumentation to perform out-reactor frequency and temperature sweep tests. This paper compares the benefits and disadvantages of various vibration instruments and summarizes the rationale behind the selection of instruments used for vibration measurements over a range of temperature and pressure pulsation frequencies. The conclusions are presented from the bench tests performed, which confirm the use of the selected instruments. (author)

  16. A subchannel and CFD analysis of void distribution for the BWR fuel bundle test benchmark

    In, Wang-Kee; Hwang, Dae-Hyun [Korea Atomic Energy Research Institute (KAERI), 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Jeong, Jae Jun, E-mail: jjjeong@pusan.ac.kr [School of Mechanical Engineering, Pusan National University, Jangjeon-dong, Geumjeong-gu, Busan 609-735 (Korea, Republic of)

    2013-05-15

    Highlights: ► We analyzed subchannel void distributions using subchannel, system and CFD codes. ► The mean error and standard deviation at steady states were compared. ► The deviation of the CFD simulation was greater than those of the others. ► The large deviation of the CFD prediction is due to interface model uncertainties. -- Abstract: The subchannel grade and microscopic void distributions in the NUPEC (Nuclear Power Engineering Corporation) BFBT (BWR Full-Size Fine-Mesh Bundle Tests) facility have been evaluated with a subchannel analysis code MATRA, a system code MARS and a CFD code CFX-10. Sixteen test series from five different test bundles were selected for the analysis of the steady-state subchannel void distributions. Four test cases for a high burn-up 8 × 8 fuel bundle with a single water rod were simulated using CFX-10 for the microscopic void distribution benchmark. Two transient cases, a turbine trip without a bypass as a typical power transient and a re-circulation pump trip as a flow transient, were also chosen for this analysis. It was found that the steady-state void distributions calculated by both the MATRA and MARS codes coincided well with the measured data in the range of thermodynamic qualities from 5 to 25%. The results of the transient calculations were also similar to each other and very reasonable. The CFD simulation reproduced the overall radial void distribution trend which produces less vapor in the central part of the bundle and more vapor in the periphery. However, the predicted variation of the void distribution inside the subchannels is small, while the measured one is large showing a very high concentration in the center of the subchannels. The variations of the void distribution between the center of the subchannels and the subchannel gap are estimated to be about 5–10% for the CFD prediction and more than 20% for the experiment.

  17. Assessment of RELAP/MOD3 Against CCFL tests with full-scale fuel bundle structures

    The purpose of this work is to investigate with the RELAP5/MOD3 (v5m5) code the influence of the structure of the core upper tie plate in a pressurized water reactor on penetration of emergency core cooling system water downwards into the core in the event of a hypothetical loss-of-coolant accident. Stationary air/water countercurrent flow experiments at atmospheric pressure for fuel bundle top area structures of the pressurized water reactors VVER-1000 and VVER-440 were simulated with the RELAP5/MOD3 (v5m5) code both without and with a countercurrent flow limitation (CCFL) correlation. The effects of flow channel size and presence of the unheated fuel rod bundle on countercurrent flow behaviour were observed. Applying the CCFL model has a minor effect on the CCFL curve in the case of our stationary calculations. A comparison with the countercurrent flow limitation in a free-flow channel is also made. The calculational results for the flow channel of a small cross-sectional area show a good agreement with the experimental results. The form of the CCFL correlation has a minor effect on the CCFL curve. (orig.) (3 refs., 47 figs., 14 tabs.)

  18. An assessment of entrainment correlations for the dryout prediction in BWR fuel bundles

    Thermal-hydraulic analysis in BWR fuel bundles usually includes calculations of detailed annular flow characteristics up to the point of dryout. State-of-the-art methods numerically resolve the governing balance equations for the relevant fields (i.e. droplet, liquid film and steam) for the system and geometry of interest (e.g. a BWR fuel bundle). However, constitutive relations are needed to close the system of equations and are fundamental to an accurate solution. One of the most important constitutive relations to consider is the droplet entrainment rate from the annular liquid film, which has an integrated effect upon the film flowrate axial distribution from the onset of annular flow (thick film) up to the dryout location (very thin film). However, currently available entrainment correlations are often developed for a relatively limit range of experimental conditions, which may not fully cover the range of applications. In this paper, we present a collection of publicly available droplet entrainment rate measurements (more than 1000 points) that have been stored into an electronic format and is used to assess the performance of several published entrainment correlations. Even though large scatter was observed for all 6 tested correlations, the model developed by Okawa et al. was shown to yield the best overall performance. (author)

  19. Spent fuel bundle counter sequence error manual - KANUPP (125 MW) NGS

    The Spent Fuel Bundle Counter (SFBC) is used to count the number and type of spent fuel transfers that occur into or out of controlled areas at CANDU reactor sites. However if the transfers are executed in a non-standard manner or the SFBC is malfunctioning, the transfers are recorded as sequence errors. Each sequence error message may contain adequate information to determine the cause of the message. This manual provides a guide to interpret the various sequence error messages that can occur and suggests probable cause or causes of the sequence errors. Each likely sequence error is presented on a 'card' in Appendix A. Note that it would be impractical to generate a sequence error card file with entries for all possible combinations of faults. Therefore the card file contains sequences with only one fault at a time. Some exceptions have been included however where experience has indicated that several faults can occur simultaneously

  20. Spent fuel bundle counter sequence error manual - RAPPS (200 MW) NGS

    The Spent Fuel Bundle Counter (SFBC) is used to count the number and type of spent fuel transfers that occur into or out of controlled areas at CANDU reactor sites. However if the transfers are executed in a non-standard manner or the SFBC is malfunctioning, the transfers are recorded as sequence errors. Each sequence error message typically contains adequate information to determine the cause of the message. This manual provides a guide to interpret the various sequence error messages that can occur and suggests probable cause or causes of the sequence errors. Each likely sequence error is presented on a 'card' in Appendix A. Note that it would be impractical to generate a sequence error card file with entries for all possible combinations of faults. Therefore the card file contains sequences with only one fault at a time. Some exceptions have been included however where experience has indicated that several faults can occur simultaneously

  1. Combustor having mixing tube bundle with baffle arrangement for directing fuel

    Hughes, Michael John; McConnaughhay, Johnie Franklin

    2016-08-23

    A combustor includes a tube bundle that extends radially across at least a portion of the combustor. The tube bundle includes an upstream surface axially separated from a downstream surface, and a plurality of tubes extend from the upstream surface through the downstream surface to provide fluid communication through the tube bundle. A barrier extends radially inside the tube bundle between the upstream and downstream surfaces, and a baffle extends axially inside the tube bundle between the upstream surface and the barrier.

  2. In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

    The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made

  3. Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

    Jayalal, M.L., E-mail: jayalal@igcar.gov.in [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Ramachandran, Suja [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Rathakrishnan, S. [Reactor Physics Section, Madras Atomic Power Station (MAPS), Kalpakkam, Tamil Nadu (India); Satya Murty, S.A.V. [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Sai Baba, M. [Resources Management Group (RMG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India)

    2015-01-15

    Highlights: • We study and compare Genetic Algorithms (GA) in the fuel bundle burnup optimization of an Indian Pressurized Heavy Water Reactor (PHWR) of 220 MWe. • Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are considered. • For the selected problem, Multi Objective GA performs better than Penalty Functions based GA. • In the present study, Multi Objective GA outperforms Penalty Functions based GA in convergence speed and better diversity in solutions. - Abstract: The work carried out as a part of application and comparison of GA techniques in nuclear reactor environment is presented in the study. The nuclear fuel management optimization problem selected for the study aims at arriving appropriate reference discharge burnup values for the two burnup zones of 220 MWe Pressurized Heavy Water Reactor (PHWR) core. Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are applied in this study. The study reveals, for the selected problem of PHWR fuel bundle burnup optimization, Multi Objective GA is more suitable than Penalty Functions based GA in the two aspects considered: by way of producing diverse feasible solutions and the convergence speed being better, i.e. it is capable of generating more number of feasible solutions, from earlier generations. It is observed that for the selected problem, the Multi Objective GA is 25.0% faster than Penalty Functions based GA with respect to CPU time, for generating 80% of the population with feasible solutions. When average computational time of fixed generations are considered, Penalty Functions based GA is 44.5% faster than Multi Objective GA. In the overall performance, the convergence speed of Multi Objective GA surpasses the computational time advantage of Penalty Functions based GA. The ability of Multi Objective GA in producing more diverse feasible solutions is a desired feature of the problem selected, that helps the

  4. Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

    Highlights: • We study and compare Genetic Algorithms (GA) in the fuel bundle burnup optimization of an Indian Pressurized Heavy Water Reactor (PHWR) of 220 MWe. • Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are considered. • For the selected problem, Multi Objective GA performs better than Penalty Functions based GA. • In the present study, Multi Objective GA outperforms Penalty Functions based GA in convergence speed and better diversity in solutions. - Abstract: The work carried out as a part of application and comparison of GA techniques in nuclear reactor environment is presented in the study. The nuclear fuel management optimization problem selected for the study aims at arriving appropriate reference discharge burnup values for the two burnup zones of 220 MWe Pressurized Heavy Water Reactor (PHWR) core. Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are applied in this study. The study reveals, for the selected problem of PHWR fuel bundle burnup optimization, Multi Objective GA is more suitable than Penalty Functions based GA in the two aspects considered: by way of producing diverse feasible solutions and the convergence speed being better, i.e. it is capable of generating more number of feasible solutions, from earlier generations. It is observed that for the selected problem, the Multi Objective GA is 25.0% faster than Penalty Functions based GA with respect to CPU time, for generating 80% of the population with feasible solutions. When average computational time of fixed generations are considered, Penalty Functions based GA is 44.5% faster than Multi Objective GA. In the overall performance, the convergence speed of Multi Objective GA surpasses the computational time advantage of Penalty Functions based GA. The ability of Multi Objective GA in producing more diverse feasible solutions is a desired feature of the problem selected, that helps the

  5. Technology Development of Integrity Evaluation of Fuel Bundles and Fuel Channel in a Two-phase Flow CANDU-6 Fuel Channel

    Two phase flow induces dynamic fluid force that causes structural vibration. Enormous vibration may result in failures of components due to the fretting wear and the fatigue, which increases the maintenance cost of the plant. From this consideration, KINS required that fuel bundles and fuel channels be evaluated to assure their integrities in high flow of more than 24 kg/s and two phase condition. Because out-reactor test loop for the simulation of two phase high flow is not available, the Wolsong CANDU-6 reactor which is in operation was utilized for the test. In-bay inspection system for the under water inspection and measurement of irradiated fuel was developed. 36 fresh fuels were measured prior to the irradiation and loaded in the fuel channel. Besides, improved method for early detection and evaluation of defect fuel was suggested

  6. Posttest examination of the VVER-1000 fuel rod bundle CORA-W2

    Sepold, L. [ed.

    1995-06-01

    The bundle meltdown experiment CORA-W2, representing the behavior of a Russian type VVER-1000 fuel element, with one B{sub 4}C/stainless steel absorber rod was selected by the OECD/CSNI as International Standard Problem (ISP-36). The experimental results of CORA-W2 serve as data base for comparison with analytical predictions of the high-temperature material behavior by various code systems. The first part of the experimental results is described in KfK 5363 (1994), the second part is documented in this report which contains the destructive post-test examination results. The metallographical and analytical (SEM/EDX) post-test examinations were performed in Germany and Russia and are summarized in five individual contributions. The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B{sub 4}C absorber rod and the stainless steel grid spacers on the ``low-temperature`` bundle damage initiation and progression. The B{sub 4}C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occurred. (orig.) [Deutsch] Das Buendel-Abschmelz-Experiment CORA-W2, das ein russisches Brennelement vom Typ WWER-1000 repraesentiert und somit auch mit einem Absorberstab aus Borkarbid/rostfreier Stahl versehen war, wurde als sog. Internationales Standardproblem (ISP-36) der OECD/CSNI ausgewaehlt. Die Versuchsergebnisse des Buendels CORA-W2 dienen als Datenbasis fuer den Vergleich mit Rechnungen mittels verschiedener Rechenprogramme im Hinblick auf das Materialverhalten bei hoher Temperatur. Der erste Teil der experimentellen Ergebnisse liegt als KfK-Bericht 5363 (1994) vor. Den zweiten Teil stellt dieser Bericht dar

  7. Advances in the manufacture of clad tubes and components for PHWR fuel bundle

    Fuel bundles for Pressurized Heavy Water Reactors (PHWRs) consists of Uranium di-oxide pellets encapsulated into thin wall Zircaloy clad tubes. Other components such as end caps, bearing pads and spacer pads are the integral elements of the fuel bundle. As the fuel assembly is subjected to severe operating conditions of high temperature and pressure in addition to continual irradiation exposure, all the components are manufactured conforming to stringent specifications with respect to chemical composition, mechanical & metallurgical properties and dimensional tolerances. The integrity of each component is ensured by NDE at different stages of manufacture. The manufacturing route for fuel tubes and components comprise of a combination of thermomechanical processing and each process step has marked effect on the final properties. The fuel tubes are manufactured by processing the extruded blanks in four stage cold pilgering with intermediate annealing and final stress relieving operation. The bar material is produced by hot extrusion followed by multi-pass swaging and intermediate annealing. Spacer pads and bearing pads are manufactured by blanking and coining of Zircaloy sheet which is made by a combination of hot and cold rolling operations. Due to the small size and stringent dimensional requirements of these appendages, selection of production route and optimization of process parameters are important. This paper discusses about various measures taken for improving the recoveries and mechanical and corrosion properties of the tube, sheet and bar materials being manufactured at Nuclear Fuel Complex, Hyderabad For the production of clad tubes, modifications at extrusion stage to reduce the wall thickness variation, introduction of ultrasonic testing of extruded blanks, optimization of cold working and heat treatment parameters at various stages of production etc. were done. The finished bar material is subjected to 100% Ultrasonic and eddy current testing to ensure

  8. Spectral element code development for incompressible flow simulations In the subchannel of a fuel rod bundle

    Two decades ago spectral element methods were developed in order to unite the the geometrical flexibility of finite element methods and the spectral convergence property of spectral methods. A code based on spectral element methods is a promising candidate to simulate turbulent incompressible fluid flow in arbitrary geometry. The aim of this work is to develop an accurate Navier-Stokes solver which is capable of simulate turbulent incompressible fluid flow in an arbitrary complex geometry. We present the concept of the spectral element methods and the algorithm used to solve Navier-Stokes equations. The design and implementation issues of a parallel spectral element code able to simulate fluid flows in arbitrary geometry are also discussed. Some preliminary results of flow simulations of in a subchannel of fuel rod bundle are presented (Authors)

  9. Numerical visualization of boiling two-phase flow behavior in fuel bundles at simulated earthquake condition

    In order to evaluate an influence of earthquake acceleration to the boiling two-phase flow behavior in nuclear reactors, numerical simulations were performed under the simulated earthquake condition. The two-phase flow analysis code, ACE-3D, was modified as the influence of the earth quake acceleration can calculate. To check out if the modification is adequate, a series of calculations were carried out and the following summaries were derived; 1) the void fraction in the fuel bundle receives the influence of the earthquake, 2) the liquid-phase in the two-phase flow moves in the same direction as the direction of oscillation due to the inputted earthquake acceleration, and 3) due to the density difference in comparison with the liquid phase, the gas phase of that moves in the direction opposite to the oscillating direction. This study enabled visualized evaluation of the boiling two-phase flow behavior in the nuclear reactors at the earthquake condition. (author)

  10. Methodology of study of the boiling crisis in a nuclear fuel rod bundle

    The boiling crisis is one of the phenomena limiting the available power from a nuclear power plant. It has been widely studied for decades, and numerous data, models, correlations or tables are now available in the literature. If we now try to obtain a general view of previous work in this field, we may note that there are several ways of tackling the subject. The mechanistic models try to modelize the two-phase flow topology and the interaction between different sublayers, and must be validated by comparison with basic experiments, such as DEBORA, where we try to get some detailed informations on the two-phase flow pattern in a pure and simple geometry. This allows us to obtain a better knowledge of the so-called 'intrinsic effect'. Up to now, these models are not yet acceptable for a nuclear use. As the geometry of the rod bundles and grids has a tremendous importance for the actual Critical Heat Flux (CHF), it is compulsory to have more precise results for a given fuel rod bundle in a restricted range of parameter: this leads to the empirical approach, using empirical CHF predictors (tables, correlations, splines, ...). One of the key points of such a method is the obtention of the local thermohydraulic values, that is to say the evaluation of the so-called 'mixing effect'. This is done by a subchannel analysis code or equivalent, which can be qualified on two kinds of experiments: overall flow measurements in a subchannel, such as HYDROMEL in single-phase flow or GRAZIELLA in two-phase flow, or detailed measurements inside a subchannel, such as AGATE. Nevertheless, the final qualification of a specific nuclear fuel, i.e. the synthesis of these mechanistic and empirical approaches, intrinsic and mixing effects, ..., must be achieved on a global test such as OMEGA. This is the strategy used in France by CEA and his partners FRAMATOME and EdF. (author)

  11. Temperature escalation in PWR fuel rod simulator bundles due to the zircaloy/steam reaction: Test ESBU-1

    This report describes the test conduct and results of the bundle test ESBU-1. The test objective was the investigation of temperature escalation of zircaloy clad fuel rods. The investigation of the temperature escalation is part of a program of out-of-pile experiments, performed within the framework of the PNS Several Fuel Damage Program. The bundle was composed of a 3x3 array of fuel rod simulators surrounded by a zircaloy shroud which was insulated with a ZrO2 fiber ceramic wrap. The fuel rod simulators comprised a tungsten heater, UO2 annular pellets, and zircaloy cladding over a 0.4 m heated length. A steam flow of 1 g/s was inlet to the bundle. The most pronounced temperature escalation was found on the central rod. The initial heatup rate of 20C/s at 11000C increased to approximately 60C/s. The maximum temperature reached was 22500C. The following fast temperature decrease was caused by runoff of molten zircaloy. Molten zircaloy swept down the thin cladding oxide layer formed during heatup. The melt dissolved the surface of the UO2 pellets and refroze as a coherent lump in the lower part of the bundle. The remaining pellets fragmented during cooldown and formed a powdery layer on the refrozen lump. The lump was sectioned posttest at several elevations: Dissolution of UO2 by the molten zircaloy, interaction between the melt and previously oxidized zircaloy, and oxidation of the melt had occurred. (orig.)

  12. CFD analysis of multiphase coolant flow through fuel rod bundles in advanced pressure tube nuclear reactors

    The key component of a pressure tube nuclear reactor core is pressure tube filled with a stream of fuel bundles. This feature makes them suitable for CFD thermal-hydraulic analysis. A methodology for CFD analysis applied to pressure tube nuclear reactors is presented in this paper, which is focused on advanced pressure tube nuclear reactors. The complex flow conditions inside pressure tube are analysed by using the Eulerian multiphase model implemented in FLUENT CFD computer code. Fuel rods in these channels are superheated but the liquid is under high pressure, so it is sub-cooled in normal operating conditions on most of pressure tube length. In the second half of pressure tube length, the onset of boiling occurs, so the flow consists of a gas liquid mixture, with the volume of gas increasing along the length of the channel in the direction of the flow. Limited computer resources enforced us to use CFD analysis for segments of pressure tube. Significant local geometries (junctions, spacers) were simulated. Main results of this work are: prediction of main thermal-hydraulic parameters along pressure tube including CHF evaluation through fuel assemblies. (authors)

  13. Neutronic calculations regarding the new LEU 6 x 6 fuel bundle for 14 MW TRIGA - SSR, in order to increase the reactor power up to 21 MW

    Iorgulis, C.; Ciocanescu, M.; Preda, M.; Mladin, M. [Institute of Nuclear Research, Pitesti (Romania)

    1998-07-01

    In order to meet the increasing demands of terminal flux for the experimental devices which will be loaded with CANDU natural uranium pins (or clusters), is necessary to rise the reactor power up to 21 MW. In this respect we consider in our evaluations a new 6x6 TRIGA fuel bundle geometry (the actual fuel bundle contains 5x5 pins). This paper will contain a comparative analysis regarding: flux and power distribution across the 29 fuel bundles standard core, and managements patters, in order to maximize the discharge fuel burnup and core lifetime. (author)

  14. Critical power analysis with mechanistic models for nuclear fuel bundles, (1). Models and verifications for boiling water reactor application

    The critical power analysis code for BWR fuel bundles, 'CAPE-BWR', was developed. The objective of the development is to predict dryout phenomena of liquid film on fuel rod surfaces without tuning any parameters even for fuel bundle design improvements. The major features of the code are modular structure with mechanistic models and parallel computation. The calculation methods were divided into three steps: subchannel, liquid film flow and spacer effect analyses. The code was validated by the rod bundle test analyses. The overall comparison of calculated critical power with 166 measured data points showed-0.3% average difference with the standard deviation of 6.3%. The spatial domain decomposition method was applied for parallel computation of the spacer effect analysis. The parallelization efficiency was about 80%. The calculated dryout location agreed well with the measured one at the full-scale 8 x 8 bundle test. The code could trace the tendencies of the critical power depending on power distribution, spacer geometry and fluid conditions within a practical range of difference. From the calculation, difference of the critical power due to the spacer geometry was clarified to be caused by the difference of droplet deposition characteristics onto the liquid film. (author)

  15. Thermal-hydraulic design calculations for the annular fuel element with replaceable test bundles (TOAST) on the test zone position 205 of KNK II/3

    Annular fuel elements are foreseen in KNK II as carrier elements for irradiation inserts and test bundles. For the third core a reloadable annular element on position 205 is foreseen, in which replaceable 19-pin test bundles (TOAST) shall be irradiated. The present report deals with the thermal-hydraulic design of the annular carrier element and the test bundle, whereby the test bundle required additional optimization. The code CIA has been used for the calculations. Start of irradiation of the subassembly is planned at the beginning of the third core operation. After optimization of the pin-spacer geometry in the test bundle, design calculations for both bundles were performed, whereby thermal coupling between both was taken into account. The calculated mass-flows and temperature distributions are given for the nominal and the eccentric element configuration. The calculated bundle pressure losses have been corrected according to experimental results

  16. Fuel composition optimization in a 78-element fuel bundle for use in a pressure tube type supercritical water-cooled reactor

    A 78-element fuel bundle containing a plutonium-thorium fuel mixture has been proposed for a Generation IV pressure tube type supercritical water-cooled reactor. In this work, using a lattice cell model created with the code DRAGON,the lattice pitch, fuel composition (fraction of PuO2 in ThO2) and radial enrichment profile of the 78-element bundle is optimized using a merit function and a metaheuristic search algorithm.The merit function is designed such that the optimal fuel maximizes fuel utilization while minimizing peak element ratings and coolant void reactivity. A radial enrichment profile of 10 wt%, 11 wt% and 20 wt% PuO2 (inner to outer ring) with a lattice pitch of 25.0 cm was found to provide the optimal merit score based on the aforementioned criteria. (author)

  17. FEED 1.6: modelling of hydrogen diffusion and precipitation in fuel bundle zircaloy components

    An as-fabricated Zircaloy component in a CANDU® fuel bundle has certain amount of hydrogen. In addition, the Zircaloy component pickups hydrogen during operation, where sheath oxidation occurs on the water side. Hydrogen content in the Zircaloy component will change due to the diffusion under gradients of concentration and temperature. A hydrostatic stress gradient may also have some effect on hydrogen diffusion. When the local concentration of hydrogen exceeds the terminal solid solubility (TSS), hydrides will start to form (i.e., hydride precipitation). Because hydrides have a negative effect on material properties (e.g., lower ductility), the hydrogen content in Zircaloy sheath needs to be limited to ensure that the sheath strength is not affected. The FEED (Finite Element Estimate for Diffusion) code was developed to predict the local hydrogen concentration and formation of hydride. The FEED 1.6 code has the following capabilities: Model transient Hydrogen/Deuterium (H/D) diffusion in Zircaloy components (e.g., fuel sheath, endcap and endcap weld); Model H/D pickup in Zircaloy sheath; Account for the effect of gradients of concentration, temperature and stress; and, Model transient hydride precipitation and re-dissolutions. This paper describes the FEED 1.6 code, including theory, models, and some validation examples. (author)

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

    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)

  19. Integrated Planar Solid Oxide Fuel Cell: Steady-State Model of a Bundle and Validation through Single Tube Experimental Data

    Paola Costamagna

    2015-11-01

    Full Text Available This work focuses on a steady-state model developed for an integrated planar solid oxide fuel cell (IP-SOFC bundle. In this geometry, several single IP-SOFCs are deposited on a tube and electrically connected in series through interconnections. Then, several tubes are coupled to one another to form a full-sized bundle. A previously-developed and validated electrochemical model is the basis for the development of the tube model, taking into account in detail the presence of active cells, interconnections and dead areas. Mass and energy balance equations are written for the IP-SOFC tube, in the classical form adopted for chemical reactors. Based on the single tube model, a bundle model is developed. Model validation is presented based on single tube current-voltage (I-V experimental data obtained in a wide range of experimental conditions, i.e., at different temperatures and for different H2/CO/CO2/CH4/H2O/N2 mixtures as the fuel feedstock. The error of the simulation results versus I-V experimental data is less than 1% in most cases, and it grows to a value of 8% only in one case, which is discussed in detail. Finally, we report model predictions of the current density distribution and temperature distribution in a bundle, the latter being a key aspect in view of the mechanical integrity of the IP-SOFC structure.

  20. Thermal-hydraulic analysis of flow blockage in a supercritical water-cooled fuel bundle with sub-channel code

    Highlights: • COBTA-SC code shows good suitability for the blockage analysis of SCWR fuel bundle. • Several thermal-hydraulic models are incorporated and evaluated for the flow blockage of SCWR-FQT bundle. • The axial/circumferential heat conduction of fuel and heat transfer correlation are identified as the important models. • The peak cladding temperature can be reduced effectively by the safety measures of SCWR-FQT. - Abstract: Sub-channel code is nowadays the most applied method for safety analysis and thermal-hydraulic simulation of fuel assembly. It plays an indispensable role to predict the detail thermal-hydraulic behavior of the supercritical water-cooled reactor (SCWR) fuel assembly because of the strong non-uniformity within the fuel bundle. Since the coolant shows a strong variation of physical thermal property near the pseudo critical line, the local blockage in an assembly of a SCWR is of importance to safety analysis. Due to the low specific heat of supercritical water with high temperature, the blockage and the subsequent flow reduction at the downstream of the blockage will yield particular high cladding temperature. To analyze the local thermal-hydraulic parameters in the supercritical water reactor-fuel qualification test (SCWR-FQT) fuel bundle with a flow blockage caused by detachment of the wire wrap, the sub-channel code COBRA-SC is unitized. The code is validated by some blockage experiments, and it reveals a good feasibility and accuracy for the SCWR and blockage flow analysis. Some new models, e.g. the axial and circumferential heat conduction model, turbulent mixing models, pressure friction models and heat transfer correlations, are incorporated in COBRA-SC code. And their influence on the cladding temperature and mass flow distribution are evaluated and discussed. Based on the results, the appropriate models for description of the flow blockage phenomenon in SCWR assembly is identified and recommended. A transient analysis of the

  1. Validation study of thermal-hydraulic analysis program spiral for fuel pin bundle of sodium-cooled fast reactors

    Full text of publication follows: Japan Nuclear Cycle Development Institute (JNC) has been developing a numerical simulation system in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel subassemblies of sodium-cooled fast reactors under various operating conditions such as normal operation, transient condition or deformed geometry condition from the viewpoint of the assessment of fuel pin structure integrity. This paper describes the validation study of SPIRAL that is one component code of the numerical simulation system and contributes to detailed simulations of local flow and temperature fields in a wire-wrapped fuel pin bundle. SPIRAL is a multi-dimensional finite element method code that can treat complicated geometries like a fuel pin bundle. For numerical stabilization, one can choose Streamline Upwind Petrov Galerkin method and Balancing Tensor Diffusivity method. Semi-implicit solution scheme (fractional step method) developed by Ramaswamy is used for time integration. As the pressure equation matrix solver, ICCG or Gaussian elimination is applied. Energy conservation equations of coolant and structure are also solved and therefore temperature distributions of both coolant and fuel pins can be calculated. Several turbulence models, high/low Reynolds number isotropic/anisotropic models, were incorporated to the code. The code was parallelized using MPI for enhancing simulation efficiency. Pre-processor is also available for numerical grid generation for wire-wrapped fuel pin bundles by curvilinear coordinate system. Fundamental validity related to solving mass, momentum and energy conservation equations and applicability of turbulence models were confirmed by simulating several basic problems. As typical examples, two kinds of simulations using high Re number models, backward facing step flow and 4- fuel-pin bundle in rectangular duct, are introduced in this paper. The simulation results indicate that RNG k-ε model shows relatively

  2. Thermal hydraulic test apparatus to develop advanced BWR fuel bundles with spectral shift rods (SSR)

    An advanced water rod (WR) called the spectral shift rod (SSR), which replaces a conventional WR in a BWR fuel bundle, enhances the BWR's merit of uranium saving through the spectral shift operation. The SSR consists of an inlet hole, a wide ascending path, a narrow descending path and an outlet hole. The inlet hole locates below a lower tie plate (LTP) and the outlet hole is set above it. In the SSR, water boils by neutron and gamma-ray heating and water level is formed in the ascending path. This SSR water level can be controlled by core flow rate, which amplifies core void fraction change, resulting in the amplified spectral shift effect. Steady state and transient tests were conducted to evaluate SSR thermal-hydraulic characteristics under BWR operation condition. The several types of SSR configuration were tested, which covers SSR design in both next generation and conventional BWRs. In this paper, the test apparatus overview and measurement systems especially two phase water level measures in the SSR are presented. (author)

  3. Improvement of the computing speed of the FBR fuel pin bundle deformation analysis code 'BAMBOO'

    JNC has developed a coupled analysis system of a fuel pin bundle deformation analysis code 'BAMBOO' and a thermal hydraulics analysis code ASFRE-IV' for the purpose of evaluating the integrity of a subassembly under the BDI condition. This coupled analysis took much computation time because it needs convergent calculations to obtain numerically stationary solutions for thermal and mechanical behaviors. We improved the computation time of the BAMBOO code analysis to make the coupled analysis practicable. 'BAMBOO' is a FEM code and as such its matrix calculations consume large memory area to temporarily stores intermediate results in the solution of simultaneous linear equations. The code used the Hard Disk Drive (HDD) for the virtual memory area to save Random Access Memory (RAM) of the computer. However, the use of the HDD increased the computation time because Input/Output (I/O) processing with the HDD took much time in data accesses. We improved the code in order that it could conduct I/O processing only with the RAM in matrix calculations and run with in high-performance computers. This improvement considerably increased the CPU occupation rate during the simulation and reduced the total simulation time of the BAMBOO code to about one-seventh of that before the improvement. (author)

  4. Development of a FBR fuel bundle-duct interaction analysis code-BAMBOO. Analysis model and verification by Phenix high burn-up fuel subassemblies

    The bundle-duct interaction analysis code ''BAMBOO'' has been developed for the purpose of predicting deformation of a wire-wrapped fuel pin bundle of a fast breeder reactor (FBR). The BAMBOO code calculates helical bowing and oval-distortion of all the fuel pins in a fuel subassembly. We developed deformation models in order to precisely analyze the irradiation induced deformation by the code: a model to analyze fuel pin self-bowing induced by circumferential gradient of void swelling as well as thermal expansion, and a model to analyze dispersion of the orderly arrangement of a fuel pin bundle. We made deformation analyses of high burn-up fuel subassemblies in Phenix reactor and compared the calculated results with the post irradiation examination data of these subassemblies for the verification of these models. From the comparison we confirmed that the calculated values of the oval-distortion and bowing reasonably agreed with the PIE results if these models were used in the analysis of the code. (author)

  5. Ultrasonic systems for high-accuracy thickness measurement of fuel bundle bearing pads and shield plug crimps

    The performance of two ultrasonic systems, remotely operated in high radiation environment, are presented. The first system is used to measure the bearing pad height of radioactive fuel bundles located in the irradiated fuel bays, at Darlington NGS. The system was designed and commissioned to achieve an accuracy of ± 20 μm. The repeatability of results is within ± 10 μm uniformity band. The measurements are independent of testing speed, water temperature, bundle temperature, pencil geometry. Possibilities and limitations of the UT system are also presented and some improved alternatives are proposed. The second system was developed for measuring the crimp height of shield plugs (special iron casting) at Bruce B - Mark Ill development. The accuracy of measurements is ± 50 μm, with a repeatability of ± 25 μm. The results are independent of shield plug thickness variation and ovality, crimp off-set and heavy-water temperature. (author)

  6. The 24 CANFLEX-NU bundle demonstration irradiation at Wolsong-1 generating station-bundle manufacture and QA, fuel handling aspects, flasking and shipping and pie for the irradiated fuel, and follow-up documentation

    Korea Ministry of Science and Technology(MOST) has pushed and given a financial support to a KEPRI/KAERI Joint Industrialization Program of CANFLEX-NU Fuel as one of Korea's National Nuclear Mid- and Long Term R and D Program. The Industrialization Program will be conducted for 3 years from 2000 November to efficiently utilize the CANFLEX fuel technology developed by KAERI and AECL jointly, where the KAERI's works have been conducted under the Korea's national program of the mid- and long-term nuclear R and D programs since 1992. This document is a report to guideline the following activities on the safety assessment for the 24 CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel bundle demonstration irradiation at Wolsong-1 Generating Station: 'bundle manufacture and QA', 'Fuel handling aspects such as loading fuel, de-fuelling and segregation, and visual in-bay examinations', 'Flasking and shipping', 'Post-irradiation examination', and 'Follow-up documentation to be produced'

  7. OECD/NRC PSBT Benchmark: Investigating the CATHARE2 Capability to Predict Void Fraction in PWR Fuel Bundle

    A. Del Nevo

    2012-01-01

    Full Text Available Accurate prediction of steam volume fraction and of the boiling crisis (either DNB or dryout occurrence is a key safety-relevant issue. Decades of experience have been built up both in experimental investigation and code development and qualification; however, there is still a large margin to improve and refine the modelling approaches. The qualification of the traditional methods (system codes can be further enhanced by validation against high-quality experimental data (e.g., including measurement of local parameters. One of these databases, related to the void fraction measurements, is the pressurized water reactor subchannel and bundle tests (PSBT conducted by the Nuclear Power Engineering Corporation (NUPEC in Japan. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The activity presented in the paper is connected with the improvement of current approaches by comparing system code predictions with measured data on void production in PWR-type fuel bundles. It is aimed at contributing to the validation of the numerical models of CATHARE 2 code, particularly for the prediction of void fraction distribution both at subchannel and bundle scale, for different test bundle configurations and thermal-hydraulic conditions, both in steady-state and transient conditions.

  8. Development of computational technology on heat transfer and fluid flow in a nuclear fuel bundle of advanced reactor

    The assessment of the RANS(Reynolds-Averaged Navier-Stokes) based turbulence model was conducted to establish the optimal CFD system for turbulent flow and heat transfer in reactor during the first year of the project. The RANS models used in this project are the two-equation models based on the eddy viscosity assumption and the Second-Moment Closure(SMC) models. Since the nuclear fuel assembly loaded in the nuclear reactor is a rod bundle which is square or triangular array, the predictions using the various turbulence models were compared for turbulent flow in bare square and/or triangular rod bundle and the rod bundle with the flow mixing vane. The study for the second year of the project examined the CFD model and the applicability of the CFD code for the turbulent two-phase flow. The numerical predictions of lateral distributions of void fraction, phasic velocities and turbulent kinetic energy were compared against the experimental results for upward and downward bubbly flow in a vertical tube. The boiling flows in vertical tube and rod bundle were also simulated to verify the CFD results

  9. IVO/AIR-WATER-CCFL, Air/water countercurrent flow limitation experiments with full-scale fuel bundle structures

    1 - Description of test facility: The test facility consists of a vertical flow channel with different internals. The test section was principally made of transparent acrylic material to allow visual observations. One fuel bundle top area structure of the Soviet-type pressurized water reactors VVER-1000 and VVER-440 in full scale was the principal test section. In order to get experimental data on the effects of different parameters on the CCFL behaviour, various configurations of the principal test sections were studied. Plate 1 corresponds to the perforated upper tie plate in full scale of the reactor VVER-1000 and plate 12 to the upper tie plate in full scale of the reactor VVER-440. 2 - Description of test: The procedure of the model tests consisted of establishing the air inlet flow rate and then increasing the water flow rate so that the given liquid head above the perforated plate, or above the fuel rod bundle when the flow channel provided only with the bundle was reached. After the stationary conditions maintained for a prolonged period, the injected water and air flows, and the average height of the mixture level above the perforated plate were registered. All reported air and water flow rates are average values at each test point. The distance of the water inlet from the perforated plate was 2000 mm, and the water level in the water collection chamber was kept constant. Small-size plates were tested. Also the effect of the unheated fuel rod bundle and the size of the free flow channel on the CCFL behaviour were studied

  10. Two-phase flow regime observations in a vertical hexagonal flow channel with and without a finned fuel bundle

    Previous flow regime studies have been for horizontal, vertical, and inclined pipe flow. As such, only a few studies have been performed on bundle geometries. The present paper examines the flow regimes for a vertical hexagonal flow channel with and without a finned fuel bundle. This type of a 36 finned rod hexagonal fuel bundle in parallel hexagonal flow channels is used in a MAPLE (Multi- purpose Applied Physics Lattice Experimental) type nuclear reactor. An experiment apparatus was designed consisting of the flow channel, inlet plenum and an air-water separator. The inlet plenum is used to provide a uniform mixture of air and water before entering the hexagonal flow channel. A turbine flow meter is used to determine the water flow rate. The turbine flow meter is calibrated for a low flow range and limits the measurable flow to 50 l/min. Flow pattern observation is determined by a SONY video camera, Real-Time Neutron Radiography, pressure transducer and capacitance transducer. The Sony video camera provides visual observation through a Lucite flow channel. The Real-Time Neutron Radiography system allows for flow visualization through an Aluminum flow channel. The pressure drop is correlated by the Validyne pressure transducer and the capacitance transducer provides the void fraction relationship

  11. Investigations of flow and temperature field development in bare and wire-wrapped reactor fuel pin bundles cooled by sodium

    Highlights: ► We study sodium flow and temperature development in fuel pin bundles. ► Pin diameter, number of pins, wire wrap and ligament gap are varied as parameters. ► Flow development is achieved within ∼30–40 hydraulic diameters. ► Thermal development is attained only for small pin diameter and less number of pins. ► Wire wrap and ligament gap strongly influence Nusselt number. - Abstract: Simultaneous development of liquid sodium flow and temperature fields in the heat generating pin bundles of reactor has been investigated. Development characteristics are seen to be strongly influenced by pin diameter, number of pins, helical wire-wrap, ligament gap between the last row of pins and hexcan wall and Reynolds number. Flow development is achieved within an axial length of ∼125 hydraulic diameters, for all the pin bundle configurations considered. But temperature development is attained only if the pin diameter is small or the number of pins is less. In the case of large pin diameter with more pins, temperature development could not be achieved even after a length of ∼1000 hydraulic diameters. The reason for this behavior is traced to be the weak communication among sub-channels in tightly packed bundles. It is seen that the pin Nusselt number decreases from center to periphery in a bundle. Also, if the ligament gap is narrow, the Nusselt number is large and more uniform. Flow development length is short if the Reynolds number is large and the converse is true for thermal development length. Helical wire-wrap shortens the thermal entry length and significantly enhances the global Nusselt number. But, its influence on hydrodynamic entry length is not significant

  12. Feasibility evaluation of x-ray imaging for measurement of fuel rod bowing in CFTL test bundles

    The Core Flow Test Loop (CFTL) is a high temperature, high pressure, out-of-reactor helium-circulating system. It is designed for detailed study of the thermomechanical performance, at prototypic steady-state and transient operating conditions, of electrically heated rods that simulate segments of core assemblies in the Gas-Cooled Fast Breeder reactor demonstration plant. Results are presented of a feasibility evaluation of x-ray imaging for making measurements of the displacement (bowing) of fuel rods in CFTL test bundles containing electrically heated rods. A mock-up of a representative CFTL test section consisting of a test bundle and associated piping was fabricated to assist in this evaluation

  13. SEU blending project, concept to commercial operation, Part 3: production of powder for demonstration irradiation fuel bundles

    The processes for production of Slightly Enriched Uranium (SEU) dioxide powder and Blended Dysprosium and Uranium (BDU) oxide powder that were developed at laboratory scale at Cameco Technology Development (CTD), were implemented and further optimized to supply to Zircatec Precision Industries (ZPI) the quantities required for manufacturing twenty six Low Void Reactivity (LVRF) CANFLEX fuel bundles. The production of this new fuel was a challenge for CTD and involved significant amount of work to prepare and review documentation, develop and approve new analytical procedures, and go through numerous internal reviews and audits by Bruce Power, CNSC and third parties independent consultants that verified the process and product quality. The audits were conducted by Quality Assurance specialists as well as by Human Factor Engineering experts with the objective to systematically address the role of human errors in the manufacturing of New Fuel and confirm whether or not a credible basis had been established for preventing human errors. The project team successfully passed through these audits. The project management structure that was established during the SEU and BDU blending process development, which included a cross-functional project team from several departments within Cameco, maintained its functionality when Cameco Technology Development was producing the powder for manufacturing Demonstration Irradiation fuel bundles. Special emphasis was placed on the consistency of operating steps and product quality certification, independent quality surveillance, materials segregation protocol, enhanced safety requirements, and accurate uranium accountability. (author)

  14. Some aspects on security and safety in a potential transport of a CANDU spent nuclear fuel bundle, in Romania

    Each Member States (MS) is responsible for the security and safety of radioactive material during transport, since radioactive material is most vulnerable during transport. The paper presents some aspects on security and safety related to the potential transport of a CANDU Spent Nuclear Fuel (SNF) bundle from NPP CANDU Cernavoda to INR Pitesti. The possible environmental impact and radiological consequences following a potential event during transportation is analyzed, since the protection of the people and the environment is the essential goal to be achieved. Some testing for the package to be used for transportation will be also given. (author)

  15. Some aspects on security and safety in a potential transport of a CANDU spent nuclear fuel bundle, in Romania

    Vieru, G., E-mail: gheorghe.vieru@nuclear.ro [Inst. for Nuclear Research, Pitesti (Romania)

    2010-07-01

    Each Member States (MS) is responsible for the security and safety of radioactive material during transport, since radioactive material is most vulnerable during transport. The paper presents some aspects on security and safety related to the potential transport of a CANDU Spent Nuclear Fuel (SNF) bundle from NPP CANDU Cernavoda to INR Pitesti. The possible environmental impact and radiological consequences following a potential event during transportation is analyzed, since the protection of the people and the environment is the essential goal to be achieved. Some testing for the package to be used for transportation will be also given. (author)

  16. A prediction method of the effect of radial heat flux distribution on critical heat flux in CANDU fuel bundles

    Fuel irradiation experiments to study fuel behaviors have been performed in the experimental loops of the National Research Universal (NRU) Reactor at Atomic Energy of Canada Limited (AECL) Chalk River Laboratories (CRL) in support of the development of new fuel technologies. Before initiating a fuel irradiation experiment, the experimental proposal must be approved to ensure that the test fuel strings put into the NRU loops meet safety margin requirements in critical heat flux (CHF). The fuel strings in irradiation experiments can have varying degrees of fuel enrichment and burnup, resulting in large variations in radial heat flux distribution (RFD). CHF experiments performed in Freon flow at CRL for full-scale bundle strings with a number of RFDs showed a strong effect of RFD on CHF. A prediction method was derived based on experimental CHF data to account for the RFD effect on CHF. It provides good CHF predictions for various RFDs as compared to the data. However, the range of the tested RFDs in the CHF experiments is not as wide as that required in the fuel irradiation experiments. The applicability of the prediction method needs to be examined for the RFDs beyond the range tested by the CHF experiments. The Canadian subchannel code ASSERT-PV was employed to simulate the CHF behavior for RFDs that would be encountered in fuel irradiation experiments. The CHF predictions using the derived method were compared with the ASSERT simulations. It was observed that the CHF predictions agree well with the ASSERT simulations in terms of CHF, confirming the applicability of the prediction method in fuel irradiation experiments. (author)

  17. Bundled procurement

    Chen, Yongmin; Li, Jianpei

    2015-01-01

    When procuring multiple products from competing firms, a buyer may choose separate purchase, pure bundling, or mixed bundling. We show that pure bundling will generate higher buyer surplus than both separate purchase and mixed bundling, provided that trade for each good is likely to be efficient. Pure bundling is superior because it intensifies the competition between firms by reducing their cost asymmetry. Mixed bundling is inferior because it allows firms to coordinate to ...

  18. Severe fuel damage experiments performed in the QUENCH facility with 21-rod bundles of LWR-type

    The objective of the QUENCH experimental program at the Karlsruhe Research Center is to investigate core degradation and the hydrogen source term that results from quenching/flooding an uncovered core, to examine the physical/chemical behavior of overheated fuel elements under different flooding conditions, and to create a data base for model development and improvement of severe fuel damage (SFD) code systems. The large-scale 21-rod bundle experiments conducted in the QUENCH out-of-pile facility are supported by an extensive separate-effects test program, by modeling activities as well as application and improvement of SFD code systems. International cooperations exist with institutions mainly within the European Union but e.g. also with the Russian Academy of Science (IBRAE, Moscow) and the CSARP program of the USNRC. So far, eleven experiments have been performed, two of them with B4C absorber material. Experimental parameters were: the temperature at initiation of reflood, the degree of peroxidation, the quench medium, i.e. water or steam, and its injection rate, the influence of a B4C absorber rod, the effect of steam-starved conditions before quench, the influence of air oxidation before quench, and boil-off behavior of a water-filled bundle with subsequent quenching. The paper gives an overview of the QUENCH program with its organizational structure, describes the test facility and the test matrix with selected experimental results. (author)

  19. Coupling analysis of deformation and thermal-hydraulics in a FBR fuel pin bundle using BAMBOO and ASFRE-IV Codes

    The bundle-duct interaction may occur in sodium cooled wire-wrapped FBR fuel subassemblies in high burn-up conditions. JNC has been developing a bundle deformation analysis code BAMBOO (Behavior Analysis code for Mechanical interaction of fuel Bundle under On-power Operation), a thermal hydraulics analysis code ASFRE-IV (Analysis of Sodium Flow in Reactor Elements - ver. IV) and their coupling method as a simulation system for the evaluation on the integrity of deformed FBR fuel pin bundles. In this study, the simulation system was applied to a coupling analysis of deformation and thermal-hydraulics in the fuel pin-bundle under a steady-state condition just after startup for the purpose of the verification of the simulation system. The iterative calculations of deformation and thermal-hydraulics employed in the coupling analysis provided numerically unstable solutions. From the result, it was found that improvement of the coupling algorithm of BAMBOO and ASFRE-IV is necessary to reduce numerical fluctuations and to obtain better convergence by introducing such computational technique as the optimized under-relaxation method. (author)

  20. Averaging methods of the gap heat transfer coefficients and the loss form coefficients of nuclear reactor cores loaded with different fuel bundles

    When performing transient analysis in heterogeneous nuclear reactors loaded with different types of fuel bundles is necessary to model the reactor core by a few representative fuel elements with average properties of a region containing a large number of fuel elements. The properties of these representative fuel bundles are obtained by averaging the thermal-hydraulic properties of the fuel elements contained in each region. In this paper we study the different ways to perform the averaging of the thermal-hydraulic properties that can have an influence on the transient results for licence purposes. Also we study the influence of the different averaging methods on the peak clad temperature (PCT) evolution for a LOCA, and on the critical power ratio (CPR) in the hot channels for a turbine trip transient without bypass credit.

  1. Numerical analysis on thermal-hydraulics of supercritical water flowing in a tight-lattice fuel bundle

    To evaluate thermal hydraulic characteristics of a tight-lattice fuel bundle of supercritical water reactor (Super Fast Reactor), a simplified 19-rod fuel assembly was analyzed with a three-dimensional two-fluid model analysis code ACE-3D which has been enhanced by Japan Atomic Energy Agency. In the ACE-3D, a two-phase flow turbulent model based on the k-ε model was adopted. In this calculation, a one-twelfth model is adopted as the computational domain taking advantage of symmetry. As the boundary conditions, mass velocity, inlet enthalpy and power per rod are to be the same as the steady state condition of the Super Fast Reactor. Cross-sectional local power distribution in the fuel assembly is set to be flat. Effect of grid spacers is taken into account in the analysis. Calculated rod surface temperatures take values near the top of the rods. Maximum clad surface temperature (MCST) is observed at the position facing to the narrowest gap on the center rod near the outlet and the value is 901K (628degC) that is almost the same as results without grid spacers. It was confirm that the predicted MCST satisfies a thermal design criteria to ensure fuel and cladding integrity: the MCST should be less than 650degC. (author)

  2. Thermo- and fluiddynamic analysis of the gas cooled fuel element bundles taking into account thermal radiation and thermal conduction

    A mathematical model has been developed, which performs the analysis of the thermal radiation between the walls and of the thermal conduction within pins and liner of a gas-cooled fuel element bundle. By means of a particular procedure, the model has been coupled with a flow-model. In this manner all important heat transfer phenomena in the thermo-fluiddynamic analysis of the bundle can be considered. Furthermore it will be possible to analyse the influence of the wall temperature distribution on the flow distribution. With the developed model a number of experiments have been computed, which have been performed with various rodbundles, in a wide range of Reynolds numbers (from laminar to turbulent), at different conditions of heating and with various gases as coolants. The computed results have been compared with the measured temperature-and pressure distributions, in order to check the validity of the model and to estimate the relative importance of the different heat transfer modes. (orig.)

  3. Double-D water rod for 9 by 9 fuel bundle

    This patent describes an improved fuel assembly including a lower tie-plate, an upper tie-plate, a square sectioned channel connecting the lower and upper tie-plate in fluid tight relation whereby fluid entering the lower tie-plate is discharged out the upper tie-plate, fuel rods each containing fissionable material therewithin. The fuel rods being held at the upper and lower tie-plates in a 9 by 9 array of rows and columns with all fuel rods having the same diameter; a plurality of spacers placed between the fuel rods for maintaining the fuel rods in spaced apart relation between the upper and lower tie-plates. The fuel rods in the 9 by 9 array having three the fuel rods removed from the middle row and two the fuel rods removed from each row on either side of the middle row to create a vacated interstitial volume defined by the absence of the removed fuel rods. The removal of the fuel rods at each row on either side displaced towards adjacent corners of the 9 by 9 array

  4. Simultaneous development of flow and temperature fields in wire-wrapped fuel pin bundles of sodium cooled fast reactor

    Simultaneous development of flow and temperature fields in the entrance region of fast breeder reactor (FBR) fuel pin bundles with helical spacer wires has been investigated by three-dimensional computational simulations. The Reynolds number, pitch of helical spacer wire and number of pins in the bundle are systematically varied. It is found that the magnitude of mean cross-stream velocity in the fully developed region is inversely proportional to the helical pitch length and it is nearly independent of the number of pins. But, there is a strong correlation between the locations of spacer wire and the peak cross-stream velocity. Flow attains full development at an axial length of 70 times hydraulic diameter in all the cases and this length is found to be unaffected by the helical pitch length. Friction factor is seen to fluctuate periodically over a mean value and the fluctuation over each helical pitch corresponds to a specific position of helical wire. The mean value of the friction factor in the entrance region reduces below the mean value in the fully developed region contrary to that seen in ducted flows. The mean fully developed friction factor is inversely proportional to the helical pitch. But, it is independent of the number of pins in the bundle. The Nusselt number passes through multiple minima before attaining fully developed periodic fluctuations and its development is slower than that of friction factor. For larger number of pins thermal development length is larger. Traditionally, the correlations reported for fully developed flow are considered for core design. But, the present study indicates that this approach is not conservative. Further, the entrance region effects and the oscillations in the fully developed region have to be properly accounted in the core design. Nusselt number exhibits a strong dependence on helical pitch similar to that of friction factor. A correlation for Nusselt number is proposed as a function of helical pitch and other

  5. CFD activities in support of thermal-hydraulic modeling of SFR fuel bundles

    Extensive testing and validation work is being performed to assess and validate Computational Fluid Dynamics (CFD) applicability to the simulation of SFR fuel assemblies. The demonstrated robustness of the method allows extending the CFD analysis to distorted fuel configurations, which will inevitably occur during extended fuel operation. The subchannel code COBRA-IV-I-MIT is adopted to evaluate the range of applicability of lumped parameter methods. Comparisons of mixing simulations show some intrinsic limitation in the subchannel methods, but allow confirming its overall applicability to nominal and mildly deformed assembly configurations. For significantly deformed geometries CFD is the recommend approach and is applied in this work. Deformed geometries considered include duct swelling, rod swelling, rod bowing, rod twisting, and various combinations of the simple deformations. While not derived from the realistic analysis of the in-core fuel behavior, the distorted geometries have been designed to embrace all conceptual worst case scenarios. The work focuses on the evaluation of the influence of the deformation on the fuel behavior, rather than on the actual fuel performance. Such approach is driven by the objective of deriving general understanding, and evaluating the applicability of subchannel analysis codes to long life fuel design, possibly in combination with distorted-channel factors derived from the CFD analyses. (author)

  6. Analysis of fuel rod behaviour within a rod bundle of a pressurized water reactor under the conditions of a loss of coolant accident (LOCA) using probabilistic methodology

    The assessment of fuel rod behaviour under PWR LOCA conditions aims at the evaluation of the peak cladding temperatures and the (final) maximum circumferential cladding strains. Moreover, the estimation of the amount of possible coolant channel blockages within a rod bundle is of special interest, as large coplanar clad strains of adjacent rods may result in strong local reductions of coolant channel areas. Coolant channel blockages of large radial extent may impair the long-term coolability of the corresponding rods. A model has been developed to describe these accident consequences using probabilistic methodology. This model is applied to study the behaviour of fuel rods under accident conditions following the double-ended pipe rupture between collant pump and pressure vessel in the primary system of a 1300 MW(el)-PWR. Specifically a rod bundle is considered consisting of 236 fuel rods, that is subjected to severe thermal and mechanical loading. The results obtained indicate that plastic clad deformations with circumferential clad strains of more than 30% cannot be excluded for hot rods of the reference bundle. However, coplanar coolant channel blockages of significant extent seem to be probable within that bundle only under certain boundary conditions which are assumed to be pessimistic. (orig./RW)

  7. Development of neural network for predicting local power distributions in BWR fuel bundles considering burnable neutron absorber

    A neural network model is under development to predict the local power distribution in a BWR fuel bundle as a high speed simulator of precise nuclear physical analysis model. The relation between 235U enrichment of fuel rods and local peaking factor (LPF) has been learned using a two-layered neural network model ENET. The training signals used were 33 patterns having considered a line symmetry of a 8x8 assembly lattice including 4 water rods. The ENET model is used in the first stage and a new model GNET which learns the change of LPFs caused by burnable neutron absorber Gadolinia, is added to the ENET in the second stage. Using this two-staged model EGNET, total number of training signals can be decreased to 99. These training signals are for zero-burnup cases. The effect of Gadolinia on LPF has a large nonlinearity and the GNET should have three layers. This combined model of EGNET can predict the training signals within 0.02 of LPF error, and the LPF of a high power rod is predictable within 0.03 error for Gadolinia rod distributions different from the training signals when the number of Gadolinia rods is less than 10. The computing speed of EGNET is more than 100 times faster than that of a precise nuclear analysis model, and EGNET is suitable for scoping survey analysis. (author)

  8. Large-scale simulations on thermal-hydraulics in fuel bundles of advanced nuclear reactors (Annual Report of the Earth Simulator Center, Dec 2008, 2007 issue)

    In order to predict the water-vapor two-phase flow dynamics in a fuel bundle of an advanced light-water reactor, large-scale numerical simulations were performed using a highly parallel-vector supercomputer, the earth simulator. Although conventional analysis methods such as subchannel codes and system analysis codes need composition equations based on the experimental data, it is difficult to obtain high prediction accuracy when experimental data to obtain the composition equations. Then, the present large-scale direct simulation method of water-vapor two-phase flow was proposed. The void fraction distribution in a fuel bundle under boiling heat transfer condition was analyzed and the bubble dynamics around the fuel rod surface were predicted quantitatively. (author)

  9. Transient non-boiling heat transfer in a fuel rod bundle during accidental power excursions

    The physical problem studied is the transient non-boiling heat transfer of a cylindrical fuel rod consisting of fuel, gap, and cladding to a steady, fully developed turbulent flow. The fuel pin is assumed to be located in the interior region of a subassembly with regular triangular or square arrangements. The turbulent velocity field as well as turbulent transport properties are specified as functions of the coordinates normal to the axial flow direction. The heat generation within the fuel may be specified as an arbitrary function of the three spatial coordinates and time. A digital computer program has been developed. On the basis of finite-difference techniques, to solve the governing partial differential equations with their associated subsidiary conditions. Results have been obtained for a series of exponential power transients of interest to safety of liquid-metal and water cooled nuclear reactors. The general physical features of transient convective heat transfer as explored by previous investigators have qualitatively been substantiated by the present analysis. Emphasis has been devoted to investigate the differences of heat-transfer (coefficient) results from multi-region analysis including a realistic fuel rod model and single-region analysis for the coolant region only. A comparison with the engineering relationships for turbulent liquid-metal cooling by Stein, which are an extension of the heat transfer coefficient concept to account for transient heat fluxes, clearly demonstrates that, at the parameters studied, Stein's approach tends to largely overestimate the convective heat transfer at early times

  10. Thermo- and fluid-dynamic studies on fuel rod and absorber bundles

    The operating safety of a nuclear reactor requires a more reliable strength analysis of the core elements subject to high stresses (fuel, breeding and absorber elements). This is among other things in a decisive way dependent on: - the maximum operating temperatures of the core element components, - the temperature gradients, - the rate of temperature variations. The calculation of these quantities as good as possible is the subject of the thermodynamic and fluid dynamic design of core elements and core. (orig.)

  11. Spectroscopic verification of fuel bundles at Embalse using CdZnTe

    The Central Nuclear Embalse is a Candu-6 nuclear power station in Argentina. In support of the International Atomic Energy Agency plan to implement remote monitoring at this site, we have developed and tested a prototype underwater spent-fuel verification system based on coplanar-grid cadmium-zinc-telluride (CdZnTe) technology. The system uses the 137 Cs gamma ray signature, and is designed for minimal interference to the operator and eventual unattended operation: Test results suggest that the method is very likely to succeed. (author)

  12. Detail design of test loop for FIV in fuel bundle and preliminary test

    Sim, Woo Gunl; Lee, Wan Young; Kim, Sung Won [Hannam University, Taejeon (Korea)

    2002-04-01

    It is urgent to develop the analytical model for the structural/mechanical integrity of fuel rod. In general, it is not easy to develop a pure analytical model. Occasionally, experimental results have been utilized for the model.Because of this reason, it is required to design proper test loop. Using the optimized test loop, With the optimized test loop, the dynamic behaviour of the rod will be evaluated and the critical flow velocity, which the rod loses the stability in, will be measured for the design of the rod. To verify the integrity of the fuel rod, it is required to evaluate the dynamic behaviour and the critical flow velocity with the test loop. The test results will be utilized to the design of the rod. Generally, the rod has a ground vibration due to turbulence in wide range of flow velocity and the amplitude of vibration becomes larger by the resonance, in a range of the velocity where occurs vortex. The rod loses stability in critical flow velocity caused by fluid-elastic instability. For the purpose of the present work to perform the conceptional design of the test loop, it is necessary (1) to understand the mechanism of the flow-induced vibration and the related experimental coefficients, (2) to evaluate the existing test loops for improving the loop with design parameters and (3) to decide the design specifications of the major equipments of the loop. 35 refs., 14 figs., 4 tabs. (Author)

  13. Large Eddy Simulation of turbulent flow in wire wrapped fuel pin bundles cooled by sodium

    The objective of the study is to understand the thermal hydraulics in a core sub-assembly with liquid sodium as coolant by performing detailed numerical simulations. The passage for the coolant flow between the fuel rods is maintained by thin wires wrapped around the rods. The contact point between the fuel pin and the spacer wire is the region of creation of hot spots and a cyclic variation of temperature in hot spots can adversely affect the mechanical properties of the clad due to the phenomena like thermal stripping. The current status quo provides two different models to perform the numerical simulations, namely Reynolds Averaged Navier-Stokes (RANS) and Large Eddy Simulation (LES). The two models differ in the extent of modelling used to close the Navier-Stokes equations. LES is a filtered approach where the large scale of motions are explicitly resolved while the small scale motions are modelled whereas RANS is a time averaging approach where all scale of motions are modelled. Thus LES involves less modelling as compared to RANS and so the results are comparatively more accurate. An attempt has been made to use the LES model. The simulations have been performed using the code Trio-U (developed by CEA). The turbulent statistics of the flow and thermal quantities are calculated. Finally the goal is to obtain the frequency of temperature oscillations at the region of hot spots near the spacer wire. (authors)

  14. The clearance potential index and hazard factors of CANDU fuel bundle and a comparison of experimental-calculated inventories

    In the field of radioactive waste management, the radiotoxicity can be characterized by two different approaches: 1) IAEA, 2004 RS-G-1.7 clearance concept and 2) US, 10CFR20 radioactivity concentration guides in terms of ingestion / inhalation hazard expressed in m3 of water/air. A comparison between the two existing safety concepts was made in the paper. The modeled case was a CANDU natural uranium, 37 elements fuel bundle with a reference burnup of 685 GJ/kgU (7928.24 MWd/tU). The radiotoxicity of the light nuclide inventories, actinide, and fission-products was calculated in the paper. The calculation was made using the ORIGEN-S from ORIGEN4.4a in conjunction with the activation-burnup library and an updated decay data library with clearance levels data in ORIGEN format produced by WIMS-AECL/SCALENEA-1 code system. Both the radioactivity concentration expressed in Curie and Becquerel, and the clearance index and ingestion / inhalation hazard were calculated for the radionuclides contained in 1 kg of irradiated fuel element at shutdown and for 1, 50, 1500 years cooling time. This study required a complex activity that consisted of various phases such us: the acquisition, setting up, validation and application of procedures, codes and libraries. For the validation phase of the study, the objective was to compare the measured inventories of selected actinide and fission products radionuclides in an element from a Pickering CANDU reactor with inventories predicted using a recent version of the ORIGEN-ARP from SCALE 5 coupled with the time dependent cross sections library, CANDU 28.lib, produced by the sequence SAS2H of SCALE 4.4a. In this way, the procedures, codes and libraries for the characterization of radioactive material in terms of radioactive inventories, clearance, and biological hazard factors are being qualified and validated, in support for the safety management of the radioactive wastes

  15. Numerical prediction of critical heat flux in nuclear fuel rod bundles with advanced three-fluid multidimensional porous media based model

    Full text of publication follows: The modern design of nuclear fuel rod bundles for Boiling Water Reactors (BWRs) is characterised with increased number of rods in the bundle, introduced part-length fuel rods and a water channel positioned along the bundle asymmetrically in regard to the centre of the bundle cross section. Such design causes significant spatial differences of volumetric heat flux, steam void fraction distribution, mass flux rate and other thermal-hydraulic parameters important for efficient cooling of nuclear fuel rods during normal steady-state and transient conditions. The prediction of the Critical Heat Flux (CHF) under these complex thermal-hydraulic conditions is of the prime importance for the safe and economic BWR operation. An efficient numerical method for the CHF prediction is developed based on the porous medium concept and multi-fluid two-phase flow models. Fuel rod bundle is observed as a porous medium with a two-phase flow through it. Coolant flow from the bundle entrance to the exit is characterised with the subsequent change of one-phase and several two-phase flow patterns. One fluid (one-phase) model is used for the prediction of liquid heating up in the bundle entrance region. Two-fluid modelling approach is applied to the bubbly and churn-turbulent vapour and liquid flows. Three-fluid modelling approach is applied to the annular flow pattern: liquid film on the rods wall, steam flow and droplets entrained in the steam stream. Every fluid stream in applied multi-fluid models is described with the mass, momentum and energy balance equations. Closure laws for the prediction of interfacial transfer processes are stated with the special emphasis on the prediction of the steam-water interface drag force, through the interface drag coefficient, and droplets entrainment and deposition rates for three-fluid annular flow model. The model implies non-equilibrium thermal and flow conditions. A new mechanistic approach for the CHF prediction

  16. Three-dimensional porous media based numerical investigation of spatial power distribution effect on advanced nuclear fuel rod bundles critical power

    The influence of spatial power generation shape on thermal-hydraulics behaviour of the fuel rod bundle has been investigated. Particularly, the occurrence of the local Boiling Transition has been analysed, indicating that conditions for the Critical Heat Flux (CHF) are reached somewhere within the boiling water channels in the assembly. The two-phase coolant flow within the bundle is represented with the two-fluid model in 3D space. The porous medium concept is applied in the simulation of the two-phase flow through the rod bundle implying nonequilibrium thermal and flow conditions. The governing equations in three-dimensions are discretized with the control volume method. The 3D numerical simulation and analyses of thermal-hydraulics in a complex geometry of an advanced nuclear fuel assembly are performed for conditions of a partial and/or complete rods uncovering indicating occurrence of high quality CHF - Dryout. The obtained results from numerical simulations are compared with experimental Critical Power data obtained from full scale tests. Employed is an electrically heated test rod bundle with real 1:1 geometry. Different radial and axial power distributions are used with wide range of inlet mass flow rates (2 - 19 kg/s) and coolant inlet subcooling (25 - 185 kJ/kg). The coolant pressure, equal to 6.9 MPa, is typical for BWRs conditions. Comparison of the predicted Critical Power values with measured data shows encouraging agreements for all analysed power distributions and the results completely reflect measured two-phase mixture cross flows, steam void distribution and spatial positions of Dryout onsets. Based on performed numerical investigation, an improvement of Dryout criteria is proposed. Dynamic effects of power shape change on spatial thermal hydraulics and hence on CHF occurrence as well as the influence of transfer function on thermal hydraulics under cyclic power and/or flow rate changes are also being analysed. Experiments for such verifications

  17. Three-dimensional porous media based numerical investigation of spatial power distribution effect on advanced nuclear fuel rod bundles critical power

    Stosic, Zoran V. [Framatome ANP GmbH . NBTT, Erlangen (Germany)], e-mail: Zoran.Stosic@Framatome-ANP.de; Stevanovic, Vladimir D. [Framatome ANP GmbH, Erlangen (Germany); Iguchi, Tadashi [Japan Atomic Energy Research Institute (JAERI), Ibaraki (Japan)

    2001-07-01

    The influence of spatial power generation shape on thermal-hydraulics behaviour of the fuel rod bundle has been investigated. Particularly, the occurrence of the local Boiling Transition has been analysed, indicating that conditions for the Critical Heat Flux (CHF) are reached somewhere within the boiling water channels in the assembly. The two-phase coolant flow within the bundle is represented with the two-fluid model in 3D space. The porous medium concept is applied in the simulation of the two-phase flow through the rod bundle implying nonequilibrium thermal and flow conditions. The governing equations in three-dimensions are discretized with the control volume method. The 3D numerical simulation and analyses of thermal-hydraulics in a complex geometry of an advanced nuclear fuel assembly are performed for conditions of a partial and/or complete rods uncovering indicating occurrence of high quality CHF - Dryout. The obtained results from numerical simulations are compared with experimental Critical Power data obtained from full scale tests. Employed is an electrically heated test rod bundle with real 1:1 geometry. Different radial and axial power distributions are used with wide range of inlet mass flow rates (2 - 19 kg/s) and coolant inlet subcooling (25 - 185 kJ/kg). The coolant pressure, equal to 6.9 MPa, is typical for BWRs conditions. Comparison of the predicted Critical Power values with measured data shows encouraging agreements for all analysed power distributions and the results completely reflect measured two-phase mixture cross flows, steam void distribution and spatial positions of Dryout onsets. Based on performed numerical investigation, an improvement of Dryout criteria is proposed. Dynamic effects of power shape change on spatial thermal hydraulics and hence on CHF occurrence as well as the influence of transfer function on thermal hydraulics under cyclic power and/or flow rate changes are also being analysed. Experiments for such verifications

  18. Influence of fuel bundle loading errors on the subcriticality during refueling campaigns for the present BWR cores of KRB-II

    On the basis of real fuel assembly inventories as they are presently available in KRB-II, the influence of fuel bundle loading errors on the subcriticality during refueling campaigns was investigated with the calculational methods of the incore fuel management. To this, control rod cells which show the least shut-down reactivity were considered and less reactive fuel assemblies were successively exchanged with fuel assemblies of highest possible reactivity from distant core regions. The results show that the total shut-down reactivity is only reduced by a comparatively small amount. The stuck rod shut-down reactivity, on the other hand, is strongly diminished with increasing number of locally concentrated mislocated fuel assemblies of highest possible reactivity. Thus, unintentional criticality cannot be reached during refueling campaigns with all control rods inserted. In conjunction with the deliberate withdrawal of one control rod, two or three mislocated fuel assemblies can cause criticality, depending on the absolute value of the realized stuck rod shut-down reactivity. (orig.)

  19. CFD simulating the transient thermal–hydraulic characteristics in a 17 × 17 bundle for a spent fuel pool under the loss of external cooling system accident

    Highlights: • A 3-D CFD is adopted to simulate transient behaviors in an SFP under the accident. • This model realistically simulates a 17 × 17 bundle, rid of porous media approach. • The loss of external cooling system accident for an SFP is assumed in this paper. • Thermal–hydraulic characteristics in a bundle are strongly influenced by grids. • The results confirm temperature rising rate used in Maanshan NPP is conservative. - Abstract: This paper develops a three-dimensional (3-D) transient computational fluid dynamics (CFD) model to simulate the thermal–hydraulic characteristics in a fuel bundle located in a spent fuel pool (SFP) under the loss of external cooling system accident. The SFP located in the Maanshan nuclear power plant (NPP) is selected herein. Without adopting the porous media approach usually used in the previous CFD works, this model uses a real-geometry simulation of a 17 × 17 fuel bundle, which can obtain the localized distributions of the flow and heat transfer during the accident. These distribution characteristics include several peaks in the axial distributions of flow, pressure, temperature, and Nusselt number (Nu) near the support grids, the non-uniform distribution of secondary flow, and the non-uniform temperature distribution due to flow mixing between rods, etc. According to the conditions adopted in the Procedure 597.1 (MNPP Plant Procedure 597.1, 2010) for the management of the loss-of-cooling event of the spent fuel pool in the Maanshan NPP, the temperature rising rate predicted by the present model can be equivalent to 1.26 K/h, which is the same order as that of 3.5 K/h in the this procedure. This result also confirms that the temperature rising rate used in the Procedure 597.1 for the Maanshan NPP is conservative. In addition, after the loss of external cooling system, there are about 44 h for the operator to repair the malfunctioning system or provide the alternative water source for the pool inventory to

  20. Interactions in Zircaloy/UO2 fuel rod bundles with Inconel spacers at temperatures above 1200deg C (posttest results of severe fuel damage experiments CORA-2 and CORA-3)

    In the CORA experiments test bundles of usually 16 electrically heated fuel rod simulators and nine unheated rods are subjected to temperature transients of a slow heatup rate in a steam environment. Thus, an accident sequence is simulated, which may develop from a small-break loss-of-coolant accident of an LWR. An aim of CORA-2, as a first test of its kind, was also to gain experience in the test conduct and posttest handling of UO2 specimens. CORA-3 was performed as a high-temperature test. The transient phases of CORA-2 and CORA-3 were initiated with a temperature ramp rate of 1 K/s. The temperature escalation due to the exothermal zircaloy(Zry)-steam reaction started at about 1000deg C, leading the bundles to maximum temperatures of 2000deg C and 2400deg C for tests CORA-2 and CORA-3, respectively. The test bundles resulted in severe oxidation and partial melting of the cladding, fuel dissolution by Zry/UO2 interaction, complete Inconel spacer destruction, and relocation of melts and fragments to lower elevations in the bundle, where extended blockages have formed. In both tests the fuel rod destruction set in together with the formation of initial melts from the Inconel/Zry interaction. The lower Zry spacer acted as a catcher for relocated material. In test CORA-2 the UO2 pellets partially disintegrated into fine particles. This powdering occurred during cooldown. There was no physical disintegration of fuel in test CORA-3. (orig./MM)

  1. Installation of an irradiated fuel bundle discharge counter at Bruce NGS-B 3 000 MW(e) CANDU power station

    Design, manufacture and installation of an irradiated fuel bundle discharge counter for the multi-unit CANDU Bruce NGS-B Generating Station involved contributions from the International Atomic Energy Agency (Agency), designers (AECL), contractors, manufacturers, utility and the regulatory agency. The installation at Bruce NGS-B was the first made by the Agency as a retrofit to a multi-unit CANDU reactor approaching its fist critical operation, where the whole project was the responsibility of the Agency and where the original design of the reactor had not had provision for the Agency equipment. The scheduling and integration of the installation into the normal activities involved in starting up a 3 000 MW(e) multi-unit generating station were successfully achieved. The Agency has demonstrated the capability and performance of the fuel discharge counter

  2. Application of Be-free Zr-based amorphous sputter coatings as a brazing filler metal in CANDU fuel bundle manufacture

    Amorphous sputter coatings of Be-free multi-component Zr-based alloys were applied as a novel brazing filler metal for Zircaloy-4 brazing. By applying the homogeneous and amorphous-structured layers coated by sputtering the crystalline targets, the highly reliable joints were obtained with the formation of predominantly grown α-Zr grains owing to a complete isothermal solidification, exhibiting high tensile and fatigue strengths as well as excellent corrosion resistance, which were comparable to those of Zircaloy-4 base metal. The present investigation showed that Be-free and Zr-based multi-component amorphous sputter coatings can offer great potential for brazing Zr alloys and manufacturing fuel rods in CANDU fuel bundle system. (author)

  3. Bundling biodiversity

    Heal, Geoffrey

    2002-01-01

    Biodiversity provides essential services to human societies. Many of these services are provided as public goods, so that they will typically be underprovided both by market mechanisms (because of the impossibility of excluding non-payers from using the services) and by government-run systems (because of the free rider problem). I suggest here that in some cases the public goods provided by biodiversity conservation can be bundled with private goods and their value to consumers captured in th...

  4. Development of multi-dimensional thermal hydraulic modeling using mixing factors for wire wrapped fuel pin bundles with inter-subassembly heat transfer in fast reactors

    Temperature distributions in fuel subassemblies of fast reactors interactively affect heat transfer from center to outer region of the core (inter-subassembly heat transfer) and cooling capability of an inter-wrapper flow, as well as maximum cladding temperature. The prediction of temperature distribution in the sub-assembly is, therefore one of the important issues for the reactor safety assessment. To treat the complex phenomena in the core, a multi-dimensional thermal hydraulic analysis is the most promising method. From the studies on the multi-dimensional thermal hydraulic modeling for the fuel sub-assemblies, the modeling have been recommended through the analysis of sodium experiments using driver subassembly test rig PLANDTL-DHX and blanket subassembly test rig CCTL-CFR. Computations of steady states experiments in the test rigs using the above modeling showed quite good agreement to the experimental data. In the present study, the use of this modeling was extended to transient analyses, and its applicability was examined. Firstly, non-dimensional parameters used to determine the mixing factors were modified from the ones based on bundle-averaged values to the ones by local values. Secondly, a new threshold function was derived and introduced to cut off the mixing factor of thermal plumes under inertia force dominant conditions. In the results of this validation, the accuracy was comparable between the modeling and the experimental instrumentation. Thus the present modeling is capable of predicting the thermal hydraulic fields of the wire wrapped fuel pin bundles with inter-subassembly heat transfer under the conditions from rated steady operations to transitions toward natural circulation decay heat removal modes. (J.P.N.)

  5. Implementation of a phenomenological DNB prediction model based on macroscale boiling flow processes in PWR fuel bundles

    Highlights: • A numerical framework was developed to mechanistically predict DNB in PWR bundles. • The DNB evaluation module was incorporated into the two-phase flow solver module. • Three-dimensional two-fluid model was the basis of two-phase flow solver module. • Liquid sublayer dryout model was adapted as CHF-triggering mechanism in DNB module. • Ability of DNB modeling approach was studied based on PSBT DNB tests in rod bundle. - Abstract: In this study, a numerical framework, comprising of a two-phase flow subchannel solver module and a Departure from Nucleate Boiling (DNB) evaluation module, was developed to mechanistically predict DNB in rod bundles of Pressurized Water Reactor (PWR). In this regard, the liquid sublayer dryout model was adapted as the Critical Heat Flux (CHF) triggering mechanism to reduce the dependency of the model on empirical correlations in the DNB evaluation module. To predict local flow boiling processes, a three-dimensional two-fluid formalism coupled with heat conduction was selected as the basic tool for the development of the two-phase flow subchannel analysis solver. Evaluation of the DNB modeling approach was performed against OECD/NRC NUPEC PWR Bundle tests (PSBT Benchmark) which supplied an extensive database for the development of truly mechanistic and consistent models for boiling transition and CHF. The results of the analyses demonstrated the need for additional assessment of the subcooled boiling model and the bulk condensation model implemented in the two-phase flow solver module. The proposed model slightly under-predicts the DNB power in comparison with the ones obtained from steady-state benchmark measurements. However, this prediction is acceptable compared with other codes. Another point about the DNB prediction model is that it has a conservative behavior. Examination of the axial and radial position of the first detected DNB using code-to-code comparisons on the basis of PSBT data indicated that the our

  6. Transient thermal hydraulic behaviour of the fuel bundles during on-power unloading operation in the proposed 500 MWe PHWR

    One of the main objectives under design and development of fuel in water cooled nuclear reactors is to ensure fuel integrity during spent fuel handling operation. The on-power refuelling facility adopted in the Indian Pressurized Heavy Water Reactors (PHWRs) causes exposure of the irradiated fuel, during its unloading, to wide variations in its surroundings including exposure to dry gaseous environment. Detailed analyses have been carried out to assess the fuel pin temperature transients during the entire course of its passage from within the reactor to the outside surroundings to ascertain fuel integrity. The cases of normal as well as envisaged off-normal transport operations have been considered in these calculations. The forced air cooling provisions have also been worked out to mitigate the consequences of off-normal transport operation. The present paper deals briefly with the system description, method of calculations and the results obtained for the case of spent fuel handling in the proposed 500 MWe PHWR. (author)

  7. NEPTUN-III reflooding and boil-off experiments with an LWHCR fuel rod bundle simulator: experimental results and initial code assessment efforts

    The NEPTUN test facility at Wuerenlingen has been modified to enable LWHCR-representative reflooding and boil-off experiments to be carried out. Results from a first series of forced feed reflooding tests, simulating cold-leg injection, are presented for a range of values of the parameters flooding rate, rod power and initial temperature. Rewetting of the LWHCR fuel bundle simulator was found to be possible in each case. Analysis of the NEPTUN-III reflooding experiments with RELAP5/MOD2 yield discrepant results and it has been shown, in the context of calculcations of the boil-off experiments, that some LWHCR-specific models and correlations need to be developed. (author)

  8. REBEKA bundle experiments

    This report is a summary of experimental investigations describing the fuel rod behavior in the refilling and reflooding phase of a loss-of-coolant accident of a PWR. The experiments were performed with 5x5 and 7x7 rod bundles, using indirectly electrically heated fuel rod simulators of full length with original PWR-KWU-geometry, original grid spacers and Zircaloy-4-claddings (Type Biblis B). The fuel rod simulators showed a cosine shaped axial power profile in 7 steps and continuous, respectively. The results describe the influence of the different parameters such as bundle size on the maximum coolant channel blockage, that of the cooling on the size of the circumferential strain of the cladding (azimuthal temperature distribution) a cold control rod guide thimble and the flow direction (axial temperature distribution) on the resulting coolant channel blockage. The rewetting behavior of different fuel rod simulators including ballooned and burst Zircaloy claddings is discussed as well as the influence of thermocouples on the cladding temperature history and the rewetting behavior. All results prove the coolability of a PWR in the case of a LOCA. Therefore, it can be concluded that the ECC-criteria established by licensing authorities can be fulfilled. (orig./HP)

  9. Process and device for testing vertical fuel rods of water-cooled nuclear reactors, which are collected into a fuel rod bundle

    To avoid high point loads on the frame and storage pond, a holding device for the fuel element is fitted in two unoccupied frame positions of a frame. A third frame position for accommodating a fuel element to be tested is kept free between the two unoccupied frame positions. After interlocking the fuel element magnetically with the holding device, the fuel element is lifted through the latter in the vertical direction, so that a sensor can drive between the fuel ords. The individual frame position is therefore subjected to a smaller load, as the whole device and the fuel element have a lower weight than two fuel elements. (orig./HP)

  10. Evaluation on BDI of large diameter pin bundles by out-of-pile bundle compression test

    Bundle-duct interaction (BDI) in core fuel subassemblies in fast reactors (FRs) is a limiting factor for fuel burnup. Since the large diameter fuel pin is generally believed to be a measure to improve FR fuel performance, the out-of-pile bundle compression test with large diameter pins (φ8.5mm and (φ 10.4mm) was performed to evaluate BDI in these bundles. In the compression test, bundle cross-sectional images (CT images) under BDI condition were obtained by using the X-ray computer tomography. In the main study, the CT images were numerically analyzed to evaluate deformation of the large diameter pin bundle due to BDI. The CT image analysis results revealed that pin-to-duct contact did not occur when the flat-to-flat bundle compression level reached one wire diameter (BDI level of 1dw), which indicates that BDI in large diameter pin bundles was mitigated similarly to the currently used small diameter pin bundles. In addition, the mitigation mechanism for BDI, which delays initiation of pin-to-duct contact, was investigated by using the computer code analysis. The code analysis results showed that cladding oval-distortion acted as a major mitigation mechanism for BDI as in the case of small pin diameter bundles. (author)

  11. Post-test calculation of the QUENCH-17 bundle experiment with debris formation and bottom water reflood using thermal hydraulic and severe fuel damage code SOCRAT/V3

    Vasiliev, A., E-mail: vasil@ibrae.ac.ru [Nuclear Safety Institute (IBRAE), B. Tulskaya 52, 115191 Moscow (Russian Federation); Stuckert, J., E-mail: juri.stuckert@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2015-03-15

    Highlights: • Modeling of processes in porous debris regions. • Analysis of coolability of massive debris bed. • Complexity of simulation of flow regime near boiling curve. - Abstract: The thermal hydraulic and SFD (Severe Fuel Damage) best estimate computer modeling code SOCRAT/V3 was used for the post-test analysis of the QUENCH-17 experiment performed at KIT on January 2013. The objective of this test was to examine the formation of a debris bed inside the completely oxidized region of the bundle without melt formation and to investigate the coolability behavior during the reflood. The test bundle for QUENCH-17 test was intentionally changed in comparison to basic QUENCH bundles (usually 21 heated rod simulators) with the emphasis to investigate debris behavior phenomena. Only 12 periphery fuel rod simulators were heated by centerline tungsten heaters. 9 unheated fuel rod simulators were located in the inner part of the test bundle. This is why the massive porous debris formation in the inner part of the bundle was not influenced by the presence of tungsten heaters. The QUENCH-17 test conditions simulated a hypothetical scenario of nuclear power plant severe accident sequence with debris bed formation in which the overheated up to 1800 K core would be flooded from the bottom by ECCS (Emergency Core Cooling System). The QUENCH-17 test included the following phases: (1) heat-up phase (heat-up rate up to 0.25 K/s); (2) oxidation phase (the cladding temperature about 1800 K in hottest region, steam mass flow rate 2 g/s); (3) bottom flood phase (characteristic cooling time about 600 s, water mass flow rate 10 g/s). SOCRAT/V3 computer modeling code was used for calculation of basic thermal hydraulic, oxidation and thermal mechanical behavior during all phases of the experiment. The calculated results are in a good agreement with experimental data which justifies the adequacy of modeling capabilities of SOCRAT code system.

  12. Post-test calculation of the QUENCH-17 bundle experiment with debris formation and bottom water reflood using thermal hydraulic and severe fuel damage code SOCRAT/V3

    Highlights: • Modeling of processes in porous debris regions. • Analysis of coolability of massive debris bed. • Complexity of simulation of flow regime near boiling curve. - Abstract: The thermal hydraulic and SFD (Severe Fuel Damage) best estimate computer modeling code SOCRAT/V3 was used for the post-test analysis of the QUENCH-17 experiment performed at KIT on January 2013. The objective of this test was to examine the formation of a debris bed inside the completely oxidized region of the bundle without melt formation and to investigate the coolability behavior during the reflood. The test bundle for QUENCH-17 test was intentionally changed in comparison to basic QUENCH bundles (usually 21 heated rod simulators) with the emphasis to investigate debris behavior phenomena. Only 12 periphery fuel rod simulators were heated by centerline tungsten heaters. 9 unheated fuel rod simulators were located in the inner part of the test bundle. This is why the massive porous debris formation in the inner part of the bundle was not influenced by the presence of tungsten heaters. The QUENCH-17 test conditions simulated a hypothetical scenario of nuclear power plant severe accident sequence with debris bed formation in which the overheated up to 1800 K core would be flooded from the bottom by ECCS (Emergency Core Cooling System). The QUENCH-17 test included the following phases: (1) heat-up phase (heat-up rate up to 0.25 K/s); (2) oxidation phase (the cladding temperature about 1800 K in hottest region, steam mass flow rate 2 g/s); (3) bottom flood phase (characteristic cooling time about 600 s, water mass flow rate 10 g/s). SOCRAT/V3 computer modeling code was used for calculation of basic thermal hydraulic, oxidation and thermal mechanical behavior during all phases of the experiment. The calculated results are in a good agreement with experimental data which justifies the adequacy of modeling capabilities of SOCRAT code system

  13. COBRA-IV PC: A personal computer version of COBRA-IV-I for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    COBRA-IV PC is a modified version of COBRA-IV-I, adapted for use with most IBM PC and PC-compatible desktop computers. Like COBRA-IV-I, COBRA-IV PC uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV PC as the implicit solution scheme option. An explicit solution scheme is also available, allowing the calculation of severe transients involving flow reversals, recirculations, expulsions, and reentry flows, with a pressure or flow boundary condition specified. In addition, several modifications have been incorporated into COBRA-IV PC to allow the code to run on the PC. These include a reduction in the array dimensions, the removal of the dump and restart options, and the inclusion of several code modifications by Oregon State University, most notably, a critical heat flux correlation for boiling water reactor fuel and a new solution scheme for cross-flow distribution calculations. 7 refs., 8 figs., 1 tab

  14. Experimental Investigation of Coolant Mixing in WWER and PWR Reactor Fuel Bundles by Laser Optical Techniques for CFD Validation

    Non intrusive laser optical measurements have been carried out to investigate the coolant mixing in a model of the head part of a fuel assembly of a WWER reactor. The goal of this research was to investigate the coolant flow around the point based in-core thermocouple; and also provide experimental database as a validation tool for computational fluid dynamics calculations. The experiments have been carried out on a full size scale model of the head part of WWER-440/213 fuel assembly. In this paper first the previous results of the research project is summarised, when full field velocity vectors and temperature were obtained by particle image velocimetry and planar laser induced fluorescence, respectively. Then, preliminary results of the investigation of the influence of the flow in the central tube will be reported by presenting velocity measurement results. In order to have well measurable effect, extreme flow rates have been set in the central tube by applying an inner tube with controlled flow rates. Despite the extreme conditions, the influence of the central tube to the velocity field proved to be significant. Further measurement will be done for the investigation of the effect of the gaps at the spacer fixings by displacing the inner tube vertically, and also the temperature distribution will also be determined at similar geometries by laser induced fluorescence. The aim of the measurements was to establish an experimental database, as well as the validation of computational fluid dynamics calculations. (Authors)

  15. Strategic Aspects of Bundling

    The increase of bundle supply has become widespread in several sectors (for instance in telecommunications and energy fields). This paper review relates strategic aspects of bundling. The main purpose of this paper is to analyze profitability of bundling strategies according to the degree of competition and the characteristics of goods. Moreover, bundling can be used as price discrimination tool, screening device or entry barriers. In monopoly case bundling strategy is efficient to sort consumers in different categories in order to capture a maximum of surplus. However, when competition increases, the profitability on bundling strategies depends on correlation of consumers' reservations values. (author)

  16. Equivariant bundle gerbes

    Murray, Michael K; Stevenson, Danny; Vozzo, Raymond F

    2015-01-01

    We develop the theory of simplicial extensions for bundle gerbes and their characteristic classes. This formalism is used to study descent problems and equivariance for bundle gerbes. We consider in detail two examples: the basic bundle gerbe on a unitary group and a string structure for a principal bundle. We show that the basic bundle gerbe is equivariant for the conjugation action and calculate its characteristic class and that a string structure gives rise to a bundle gerbe which is equivariant for a natural action of the String 2-group.

  17. The effect of radial power profile of DUPIC bundle on CHF

    The axial and ring power profiles of DUPIC bundle are much different from those of reference 37-element fuel bundle since a DUPIC fuel bundle is re-fabricated using spent PWR fuel and 2-bundle shift refuelling scheme is proposed to CANDU-6 reactor. In case that the ring power profile of a fuel bundle is altered, the flow and enthalpy distribution of subchannels and the radial position of CHF occurrence will be changed. Similarly, the axial power profile of a fuel channel affects CHF and axial position of CHF occurrence as well as axial enthalpy, quality and pressure distribution. The ring power profile of the DUPIC bundle as increasing burnup is altered and flattened compared to 37-element bundle and each fuel bundle in a fuel channel has a different ring power profile from the other bundles at different axial position in the same fuel channel. Therefore, how to consider the burnup or ring power effect on CHF is very important to DUPIC thermalhydraulic analysis. At present study, thermalhydraulic analysis of the DUPIC bundle was performed in consideration of ring power profile effect on CHF. The subchannel enthalpy, mass flux and CHF distribution for 0 burnup to discharged burnup (18,000 MWD/THM) of DUPIC bundle were evaluated using ASSERT subchannel code. The results were compared to those of 37-element bundle and the compatability of DUPIC bundle with an existing CANDU-6 was presented in a CHF point of view

  18. Bundling in Telecommunications

    Begoña García-Mariñoso; Xavier Martinez-Giralt; Pau Olivella

    2008-01-01

    The paper offers an overview of the literature on bundling in the telecommunications sector and its application in the Spanish market. We argue that the use of bundling in the provision of services is associated to technological reasons. Therefore, there appears no need to regulate bundling activities. However, this is not to say that other related telecom markets should not be scrutinized and regulated, or that the regulator should not pay attention to other bundling-related anticompetitive ...

  19. Canadian power reactor fuel

    The following subjects are covered: the basic CANDU fuel design, the history of the bundle design, the significant differences between CANDU and LWR fuel, bundle manufacture, fissile and structural materials and coolants used in the CANDU fuel program, fuel and material behaviour, and performance under irradiation, fuel physics and management, booster rods and reactivity mechanisms, fuel procurement, organization and industry, and fuel costs. (author)

  20. International Standard problem ISP 14: behaviour of a fuel bundle simulator during a specified heatup and flooding period (Rebeka experiment): results of post-test analyses: final comparison report

    The test consisted in investigating the non-steady material behaviour of a bundle of electrically heated fuel rod simulators with respect to local fuel temperatures, cladding strain, time to burst and local strain at location of burst, together with the thermal hydraulic boundary conditions. The original aim has not been fully achievable. The applied codes for mechanical fuel behaviour largely demonstrated their capabilities for pretest predictions when certain local fluid dynamic parameters are well known to the code users. The difficulties expected with proper analysis of thermal hydraulics of the test were confirmed, caused by the coupling between pin cooling conditions, rod upper plenum calculations and the feedback to clad deformation and burst simulation

  1. CANDU fuel performance

    The paper presents a review of CANDU fuel performance including a 28-element bundle for Pickering reactors, a 37-element bundle for the Bruce and Darlington reactors, and a 37-element bundle for the CANDU-6 reactors. Special emphasis is given to the analysis of fuel defect formation and propagation and definition of fuel element operating thresholds for normal operation and accident conditions. (author)

  2. Bundling and Tying

    Nicholas Economides

    2014-01-01

    We discuss strategic ways that sellers can use tying and bundling with requirement conditions to extract consumer surplus. We analyze different types of tying and bundling creating (i) intra-product price discrimination; (ii) intra-consumer price discrimination; and (iii) inter-product price discrimination, and assess the antitrust liability that these practices may entail. We also discuss the impact on consumers and competition, as well as potential antitrust liability of bundling “incontest...

  3. Contact fiber bundles

    Lerman, Eugene

    2003-01-01

    We define contact fiber bundles and investigate conditions for the existence of contact structures on the total space of such a bundle. The results are analogous to minimal coupling in symplectic geometry. The two applications are construction of K-contact manifolds generalizing Yamazaki's fiber join construction and a cross-section theorem for contact moment maps

  4. Principal noncommutative torus bundles

    Echterhoff, Siegfried; Nest, Ryszard; Oyono-Oyono, Herve

    2008-01-01

    of bivariant K-theory (denoted RKK-theory) due to Kasparov. Using earlier results of Echterhoff and Williams, we shall give a complete classification of principal non-commutative torus bundles up to equivariant Morita equivalence. We then study these bundles as topological fibrations (forgetting the...

  5. Radical power profile effect of DUPIC bundle on critical heat flux

    The axial and ring power profiles of DUPIC bundle are much different from those of reference 37-element fuel bundle since a DUPIC fuel bundle is -re-fabricated under proliferation resistance using spent PWR fuel and 2-bundle shift refuelling scheme of DUPIC bundle is proposed to CANDU-6 reactor. In case that the ring power porfile of a fuel bundle is altered, the flow and enthalpy distribution of subchannels and the radial position of CHF occurrence will be changed. Similarly, the axial power profile of a fuel channel affects CHF, axial position of CHF occurrence, axial enthalpy, quality and pressure distribution. The ring power profile of the DUPIC bundle as increasing burnup is much altered and flattened at high burnup, compared to 37-element bundle. It caused that one fuel bundle has a different ring power profile from the other fuel bundles at the different axial positions even in the same fuel channel. Therefore, how to consider burnup or ring power effect on CHF is very important to DUPIC thermalhydraulic analysis. At present study, thermalhydraulic analysis of a DUPIC bundles was performed in order to evaluate the ring power profile effect on CHF. The subchannel enthalpy, mass flux and CHF distribution from 0 burnup to discharged burnup (18,000 MWd/tHM) of DUPIC bundle were evaluated using ASSERT-PV subchannel code. The results of DUPIC bundles were compared to those of 37-elemental bundle and the comparability of DUPIC bundle with an existing CANDU-6 was presented in a CHF point of view

  6. Annular burnout data from rod-bundle experiments

    Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident

  7. Annular burnout data from rod bundle experiments

    Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident. Level average fluid conditions within the test section were calculated using steady-state mass and energy conservation considerations for the steady-state tests and a transient, homogeneous, equilibrium computer code for the transient tests. Unlike tube dryout, burnout within a rod bundle does not necessarily occur at one distinct axial level. The location of individual rod dryout was determined by scanning rods axially and locating the position where rod superheat increased from approx. =0 to 30 K or greater. Thermocouple instrumentation within the bundle allows the location of dryout to be determined to within approximately +.5 cm for many of the tests

  8. Countercurrent flow limitation experiments with full-scale bundle structures

    Atmospheric air/water experiments for VVER-440 and VVER-1000 fuel bundle structures have been carried out at different liquid heads above the perforated tie plate. Specific attention is given to countercurrent flow limitation across perforated upper tie plates in large channel geometries. The effects of the presence of the unheated fuel rod bundle and the thickness of the perforated plate on the countercurent flow behaviour have been observed. The found non-linear countercurrent flow behavior is discussed. (orig.)

  9. COBRA-IV-I: an interim version of COBRA for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    The COBRA-IV-I computer code uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV-I as the implicit solution scheme option. In addition to these techniques, a new explicit solution scheme is now available which allows the calculation of severe transients involving flow reversals, recirculations, expulsion and reentry flows, with a pressure or flow boundary condition specified. Significant storage compaction and reduced running times have been achieved to allow the calculation of problems involving hundreds of subchannels

  10. Restrictions of stable bundles

    Balaji, V

    2011-01-01

    The Mehta-Ramanathan theorem ensures that the restriction of a stable vector bundle to a sufficiently high degree complete intersection curve is again stable. We improve the bounds for the "sufficiently high degree" and propose a possibly optimal conjecture.

  11. Numerical model for thermal and mechanical behaviour of a CANDU 37-element bundle

    Prediction of transient fuel bundle deformations is important for assessing the integrity of fuel and the surrounding structural components under different operating conditions including accidents. For numerical simulation of the interactions between fuel bundle and pressure tube, a reliable numerical bundle model is required to predict thermal and mechanical behaviour of the fuel bundle assembly under different thermal loading conditions. To ensure realistic representations of the bundle behaviour, this model must include all of the important thermal and mechanical features of the fuel bundle, such as temperature-dependent material properties, thermal viscoplastic deformation in sheath, fuel-to-sheath interactions, endplate constraints and contacts between fuel elements. In this paper, we present a finite element based numerical model for predicting macroscopic transient thermal-mechanical behaviour of a complete 37-element CANDU nuclear fuel bundle under accident conditions and demonstrate its potential for being used to investigate fuel bundle to pressure tube interaction in future nuclear safety analyses. This bundle model has been validated against available experimental and numerical solutions and applied to various simulations involving steady-state and transient loading conditions. (author)

  12. Subtleties Concerning Conformal Tractor Bundles

    Graham, C Robin

    2012-01-01

    The realization of tractor bundles as associated bundles in conformal geometry is studied. It is shown that different natural choices of principal bundle with normal Cartan connection corresponding to a given conformal manifold can give rise to topologically distinct associated tractor bundles for the same inducing representation. Consequences for homogeneous models and conformal holonomy are described. A careful presentation is made of background material concerning standard tractor bundles and equivalence between parabolic geometries and underlying structures.

  13. Inertial gas pressure and circumferential ridge sheath strains in CANFLEX-ACR fuel

    'Full text:' ACR-1000® fuel is designed to operate with an average exit burnup of up to 20 MWd/kgU. This average exit burnup is in excess of the burnup in current CANDU® designs. The increased burnup will result in higher fission product inventory. This paper reports the evaluation of fuel sheath strains and internal gas pressures during normal operation. The internal gas pressures and sheath strains are assessed using the ELESTRES computer code. Predicted strains and pressures have adequate margin to the relevant acceptance criteria.

  14. Right bundle branch block

    Bussink, Barbara E; Holst, Anders Gaarsdal; Jespersen, Lasse;

    2013-01-01

    AimsTo determine the prevalence, predictors of newly acquired, and the prognostic value of right bundle branch block (RBBB) and incomplete RBBB (IRBBB) on a resting 12-lead electrocardiogram in men and women from the general population.Methods and resultsWe followed 18 441 participants included in...... men vs. 0.5%/2.3% in women, P <0.001). Significant predictors of newly acquired RBBB were male gender, increasing age, high systolic blood pressure, and presence of IRBBB, whereas predictors of newly acquired IRBBB were male gender, increasing age, and low BMI. Right bundle branch block was associated...... with significantly increased all-cause and cardiovascular mortality in both genders with age-adjusted hazard ratios (HR) of 1.31 [95% confidence interval (CI), 1.11-1.54] and 1.87 (95% CI, 1.48-2.36) in the gender pooled analysis with little attenuation after multiple adjustment. Right bundle branch...

  15. Principal -bundles on Nodal Curves

    Usha N Bhosle

    2001-08-01

    Let be a connected semisimple affine algebraic group defined over . We study the relation between stable, semistable -bundles on a nodal curve and representations of the fundamental group of . This study is done by extending the notion of (generalized) parabolic vector bundles to principal -bundles on the desingularization of and using the correspondence between them and principal -bundles on . We give an isomorphism of the stack of generalized parabolic bundles on with a quotient stack associated to loop groups. We show that if is simple and simply connected then the Picard group of the stack of principal -bundles on is isomorphic to ⊕ , being the number of components of .

  16. An advancement in iterative solution schemes for three-dimensional, two-fluid modeling of two-phase flow in PWR fuel bundles

    Highlights: • A fully three-dimensional two-fluid model coupled with heat conduction was outlined. • Two-fluid numerical scheme capability was evaluated against NUPEC PSBT Benchmark. • GMRES, FGMRES, DQGMRES, CGNR, BCG, and TFQMR solvers were tested as iterative schemes. • Candidate Krylov solvers do not introduce deviations to the two-phase flow results. • GMRES, FGMRES, and DQGMRES have a more efficient and stable convergence performance. - Abstract: This paper outlines a fully three-dimensional two-fluid one-pressure model with a semi-implicit finite difference scheme coupled with heat conduction which can be applicable to thermal non-equilibrium two-phase flow field in subchannel geometry of Pressurized Water Reactors (PWR). The system of equations was linearized using the Newton–Raphson method and was collapsed into the pressure equations forming a system of the Poisson type. Then, two-phase flow modeling was combined with Krylov methods as advanced computing techniques to investigate the feasibility of implementing preconditioned Krylov subspace solvers as the numerical scheme to solve pressure equations. Six popular Krylov subspace solvers were considered: GMRES, FGMRES, DQGMRES, CGNR, BCG, and TFQMR combined with the block incomplete LU factorization with a dual truncation strategy (BILUT) preconditioner. These proposed iterative solvers were applied to the constructed linear pressure equations in the inner iteration in combination with the outer-Raphson iteration loop. Evaluation was performed in two stages. First, two-fluid numerical scheme capability was evaluated against OECD/NRC NUPEC PWR Bundle tests (PSBT Benchmark). The results for steady-state (PSBT) bundle show that an overall agreement can be found. At the second stage, convergency, stability, and accuracy of the proposed schemes were studied based on PSBT steady-state data through a comparison of utilized Krylov solvers and the direct inversion method as the pressure solution

  17. Modelling of transient dynamic bundle deformation using time integration scheme

    The BOW code has been examined whether its modeling capability can be extended to the simulation of interactions (i.e., fretting) between neighbouring fuel elements in a fuel bundle and between the fuel bundle and the pressure tube in a fuel channel. The current BOW code is specialized in simulating the static problems, such as the deflection of each element and interactions between neighbouring elements in a fuel bundle, and interactions between neighbouring bundles and between a bundle and the pressure tube in a fuel channel. The Wilson θ time integration scheme has been implemented in the BOW code, for the extension of its capability to modelling dynamic contact problems. As part of verification to ensure that the modification in the code functions exactly as designed, the dynamic-modelling capability of the BOW code has been applied to simple support beam cases subjected to a uniform step load at the middle of the beam. The calculation results confirmed that the modified BOW code, where the contact algorithm is implemented in the step-by-step integration manner using the Wilson θ time integration scheme, can solve the dynamic problem with unconditional convergence. This paper describes the theory and models for the new capabilities of the BOW code. (author)

  18. Hybrid bundle divertor design

    A hybrid bundle divertor design is presented that produces <0.3% magnetic ripple at the center of the plasma while providing adequate space for the coil shielding and structure for a tokamak fusion test reactor similar to the International Tokamak Reactor and the Engineering Test Facility (with R = 5 m, B = 5 T, and a /SUB wall/ = 1.5 m, in particular). This hybrid divertor consists of a set of quadrupole ''wing'' coils running tangent to the tokamak plasma on either side of a bundle divertor. The wing coils by themselves pull the edge of the plasma out 1.5 m and spread the thickness of the scrape-off layer from 0.1 to 0.7 m at the midplane. The clear aperture of the bundle divertor throat is 1.0 m high and 1.8 m wide. For maintenance or replacement, the hybrid divertor can be disassembled into three parts, with the bundle divertor part pulling straight out between toroidal field coils and the wing coils then sliding out through the same opening

  19. Dynamics of flagellar bundling

    Janssen, Pieter; Graham, Michael

    2010-11-01

    Flagella are long thin appendages of microscopic organisms used for propulsion in low-Reynolds environments. For E. coli the flagella are driven by a molecular motor, which rotates the flagella in a counter-clockwise motion (CCM). When in a forward swimming motion, all flagella bundle up. If a motor reverses rotation direction, the flagella unbundle and the cell makes a tumbling motion. When all motors turn in the same CC direction again, the flagella bundle up, and forward swimming continues. To investigate the bundling, we consider two flexible helices next to each other, as well as several flagella attached to a spherical body. Each helix is modeled as several prolate spheroids connected at the tips by springs. For hydrodynamic interactions, we consider the flagella to made up of point forces, while the finite size of the body is incorporated via Fax'en's laws. We show that synchronization occurs quickly relative to the bundling process. For flagella next to each other, the initial deflection is generated by rotlet interactions generated by the rotating helices. At longer times, simulations show the flagella only wrap once around each other, but only for flagella that are closer than about 4 helix radii. Finally, we show a run-and-tumble motion of the body with attached flagella.

  20. On framed quantum principal bundles

    Durdevic, M

    1995-01-01

    A noncommutative-geometric formalism of framed principal bundles is sketched, in a special case of quantum bundles (over quantum spaces) possessing classical structure groups. Quantum counterparts of torsion operators and Levi-Civita type connections are analyzed. A construction of a natural differential calculus on framed bundles is described. Illustrative examples are presented.

  1. CANDU bundle junction. Misalignment probability and pressure-drop correlation

    The pressure drop over the bundle junction is an important component of the pressure drop in a CANDU (Canada Deuterium Uranium) fuel channel. This component can represent from ∼ 15% for aligned bundles to ∼ 26% for rotationally misaligned bundles, and is dependent on the degree of misalignment. The geometry of the junction increases the mixing between subchannels, and hence improves the thermal performance of the bundle immediately downstream. It is therefore important to model the junction's performance adequately. This paper summarizes a study sponsored by COG (CANDU Owners Group) and an NSERC (National Science and Engineering Research Council) Industrial Research Grant, undertaken, at CRL (Chalk River Laboratories) to identify and develop a bundle-junction model for potential implementation in the ASSERT (Advanced Solution of Subchannel Equations in Reactor Thermalhydraulics) subchannel code. The work reported in this paper consists of two components of this project: an examination of the statistics of bundle misalignment, demonstrating that there are no preferred positions for the bundles and therefore all misalignment angles are equally possible; and, an empirical model for the single-phase pressure drop across the junction as a function of the misalignment angle. The second section of this paper includes a brief literature review covering the experimental, analytical and numerical studies concerning the single-phase pressure drop across bundle junctions. 32 refs., 9 figs

  2. Cold-neutron tomography of annular flow and functional spacer performance in a model of a boiling water reactor fuel rod bundle

    Highlights: → Annular flows w/wo functional spacers are investigated by cold neutron imaging. → Liquid film thickness distribution on fuel pins and on spacer vanes is measured. → The influence of the spacers on the liquid film distributions has been quantified. → The cross-sectional averaged liquid hold-up significantly affected by the spacers. → The sapers affect the fraction of the entrained liquid hold up in the gas core. - Abstract: Dryout of the coolant liquid film at the upper part of the fuel pins of a boiling water reactor (BWR) core constitutes the type of heat transfer crisis relevant for the conditions of high void fractions. It is both a safety concern and a limiting factor in the thermal power and thus for the economy of BWRs. We have investigated adiabatic, air-water annular flows in a scaled-up model of two neighboring subchannels as found in BWR fuel assemblies using cold-neutron tomography. The imaging of the double suchannel has been performed at the ICON beamline at the neutron spallation source SINQ at the Paul Scherrer Institute, Switzerland. Cold-neutron tomography is shown here to be an excellent tool for investigating air-water annular flows and the influence of functional spacers of different geometries on such flows. The high-resolution, high-contrast measurements provide the spatial distributions of the coolant liquid film thickness on the fuel pin surfaces as well as on the surfaces of the spacer vanes. The axial variations of the cross-section averaged liquid hold-up and its fraction in the gas core shows the effect of the spacers on the redistribution of the two phases.

  3. Bundles of Banach algebras

    D. A. Robbins

    1994-12-01

    Full Text Available We study bundles of Banach algebras π:A→X, where each fiber Ax=π−1({x} is a Banach algebra and X is a compact Hausdorff space. In the case where all fibers are commutative, we investigate how the Gelfand representation of the section space algebra Γ(π relates to the Gelfand representation of the fibers. In the general case, we investigate how adjoining an identity to the bundle π:A→X relates to the standard adjunction of identities to the fibers.

  4. Helices and vector bundles

    Rudakov, A N

    1990-01-01

    This volume is devoted to the use of helices as a method for studying exceptional vector bundles, an important and natural concept in algebraic geometry. The work arises out of a series of seminars organised in Moscow by A. N. Rudakov. The first article sets up the general machinery, and later ones explore its use in various contexts. As to be expected, the approach is concrete; the theory is considered for quadrics, ruled surfaces, K3 surfaces and P3(C).

  5. Bundled monocapillary optics

    Hirsch, Gregory

    2002-01-01

    A plurality of glass or metal wires are precisely etched to form the desired shape of the individual channels of the final polycapillary optic. This shape is created by carefully controlling the withdrawal speed of a group of wires from an etchant bath. The etched wires undergo a subsequent operation to create an extremely smooth surface. This surface is coated with a layer of material which is selected to maximize the reflectivity of the radiation being used. This reflective surface may be a single layer of material, or a multilayer coating for optimizing the reflectivity in a narrower wavelength interval. The collection of individual wires is assembled into a close-packed multi-wire bundle, and the wires are bonded together in a manner which preserves the close-pack configuration, irrespective of the local wire diameter. The initial wires are then removed by either a chemical etching procedure or mechanical force. In the case of chemical etching, the bundle is generally segmented by cutting a series of etching slots. Prior to removing the wire, the capillary array is typically bonded to a support substrate. The result of the process is a bundle of precisely oriented radiation-reflecting hollow channels. The capillary optic is used for efficiently collecting and redirecting the radiation from a source of radiation which could be the anode of an x-ray tube, a plasma source, the fluorescent radiation from an electron microprobe, a synchrotron radiation source, a reactor or spallation source of neutrons, or some other source.

  6. Bundling harvester; Nippukorjausharvesteri

    Koponen, K. [Eko-Log Oy, Kuopio (Finland)

    1996-12-31

    The staring point of the project was to design and construct, by taking the silvicultural point of view into account, a harvesting and processing system especially for energy-wood, containing manually driven bundling harvester, automatizing of the harvester, and automatized loading. The equipment forms an ideal method for entrepreneur`s-line harvesting. The target is to apply the system also for owner`s-line harvesting. The profitability of the system promotes the utilization of the system in both cases. The objectives of the project were: to construct a test equipment and prototypes for all the project stages, to carry out terrain and strain tests in order to examine the usability and durability, as well as the capacity of the machine, to test the applicability of the Eko-Log system in simultaneous harvesting of energy and pulp woods, and to start the marketing and manufacturing of the products. The basic problems of the construction of the bundling harvester have been solved using terrain-tests. The prototype machine has been shown to be operable. Loading of the bundles to form sufficiently economically transportable loads has been studied, and simultaneously, the branch-biomass has been tried to be utilized without loosing the profitability of transportation. The results have been promising, and will promote the profitable utilization of wood-energy

  7. Fuel assembly

    A fuel assembly of a BWR type reactor comprises a rectangular parallelopiped channel box and fuel bundles contained in the channel box. The fuel bundle comprises an upper tie plate, a lower tie plate, a plurality of spacers a plurality of fuel rods and a water rod. In each fuel rod, the amount of fission products is reduced at upper and lower end regions of an effective fuel portion than that in other regions of the effective fuel region. In a portion of the fuel rods, fuel pellets containing burnable poisons are disposed at the upper and lower end regions. In addition, the upper and lower portions are constituted with natural uranium. Each of the upper and lower end regions is not greater than 15% of the effective fuel length. Since this can enhance reactivity control effect without worsening fuel economy, the control amount for excess reactivity upon long-term cycle operation can be increased. (I.N.)

  8. Kernel Bundle EPDiff

    Sommer, Stefan Horst; Lauze, Francois Bernard; Nielsen, Mads; Pennec, Xavier

    information to be automatically incorporated in registrations and promises to improve the standard framework in several aspects. We present the mathematical foundations of LDDKBM and derive the KB-EPDiff evolution equations, which provide optimal warps in this new framework. To illustrate the resulting......In the LDDMM framework, optimal warps for image registration are found as end-points of critical paths for an energy functional, and the EPDiff equations describe the evolution along such paths. The Large Deformation Diffeomorphic Kernel Bundle Mapping (LDDKBM) extension of LDDMM allows scale space...... diffeomorphism paths, we give examples showing the decoupled evolution across scales and how the method automatically incorporates deforma- tion at appropriate scales....

  9. Twists of symmetric bundles

    Cassou-Nogues, Ph.; Erez, B.; Taylor, M. J.

    2004-01-01

    We establish comparison results between the Hasse-Witt invariants w_t(E) of a symmetric bundle E over a scheme and the invariants of one of its twists E_{\\alpha}. For general twists we describe the difference between w_t(E) and w_t(E_{\\alpha}) up to terms of degree 3. Next we consider a special kind of twist, which has been studied by A. Fr\\"ohlich. This arises from twisting by a cocycle obtained from an orthogonal representation. We show how to explicitly describe the twist for representatio...

  10. TRIGA modified bundle thermal hydraulic analysis

    Negut, G.; Mladin, M.; Preda, M. [Inst. for Nuclear Research, Pitesti (Romania)

    2001-07-01

    TRIGA 14 MW steady state reactor (SSR) has more than 20 years of operation experience. It was used as a material test reactor to accomplish full range of experiments of CANDU type fuel, tests on structure material as Zircaloy and stainless steel. We did, also, isotope production for industrial and medical use, neutronography, gamma prompt, neutron diffractometry and activation analysis. In order to optimize the core for a more homogenous burnup we did some experiments on a modified fuel bundle. The paper is dedicated to the computations done in order to validate the optimized core configuration. The analysis has shown no significant impact on the central fuel temperatures, to affect the core safety. (orig.)

  11. TRIGA modified bundle thermal hydraulic analysis

    TRIGA 14 MW steady state reactor (SSR) has more than 20 years of operation experience. It was used as a material test reactor to accomplish full range of experiments of CANDU type fuel, tests on structure material as Zircaloy and stainless steel. We did, also, isotope production for industrial and medical use, neutronography, gamma prompt, neutron diffractometry and activation analysis. In order to optimize the core for a more homogenous burnup we did some experiments on a modified fuel bundle. The paper is dedicated to the computations done in order to validate the optimized core configuration. The analysis has shown no significant impact on the central fuel temperatures, to affect the core safety. (orig.)

  12. Bundle Security Protocol for ION

    Burleigh, Scott C.; Birrane, Edward J.; Krupiarz, Christopher

    2011-01-01

    This software implements bundle authentication, conforming to the Delay-Tolerant Networking (DTN) Internet Draft on Bundle Security Protocol (BSP), for the Interplanetary Overlay Network (ION) implementation of DTN. This is the only implementation of BSP that is integrated with ION.

  13. Fiber Bundles and Parseval Frames

    Agrawal, Devanshu; Knisley, Jeff

    2015-01-01

    Continuous frames over a Hilbert space have a rich and sophisticated structure that can be represented in the form of a fiber bundle. The fiber bundle structure reveals the central importance of Parseval frames and the extent to which Parseval frames generalize the notion of an orthonormal basis.

  14. Fiber bundle phase conjugate mirror

    Ward, Benjamin G.

    2012-05-01

    An improved method and apparatus for passively conjugating the phases of a distorted wavefronts resulting from optical phase mismatch between elements of a fiber laser array are disclosed. A method for passively conjugating a distorted wavefront comprises the steps of: multiplexing a plurality of probe fibers and a bundle pump fiber in a fiber bundle array; passing the multiplexed output from the fiber bundle array through a collimating lens and into one portion of a non-linear medium; passing the output from a pump collection fiber through a focusing lens and into another portion of the non-linear medium so that the output from the pump collection fiber mixes with the multiplexed output from the fiber bundle; adjusting one or more degrees of freedom of one or more of the fiber bundle array, the collimating lens, the focusing lens, the non-linear medium, or the pump collection fiber to produce a standing wave in the non-linear medium.

  15. Twisted Vector Bundles on Pointed Nodal Curves

    Ivan Kausz

    2005-05-01

    Motivated by the quest for a good compactification of the moduli space of -bundles on a nodal curve we establish a striking relationship between Abramovich’s and Vistoli’s twisted bundles and Gieseker vector bundles.

  16. Fuel assembly for BWR-type reactor

    74 fuel rods and 2 large diameter water rods are disposed in 9 x 9 square lattice. Both upper and lower ends thereof are bundled by tie plates to constitute a fuel bundle, and the fuel bundle is surrounded by a channel box. Among eight short fuel rods, four short fuel rods are disposed to four corners on the second layer from the outermost circumference of the fuel bundle, and four short fuel rods are disposed to the center of each of the sides at the outermost circumference of the fuel bundle. Eight long fuel rods are disposed in adjacent with the short fuel rods at the outermost circumference of the fuel bundle. Eight long fuel rods are disposed to the second layer from the outermost circumference of the fuel bundle and in adjacent with the former eight long fuel rods. The long fuel rods contain burnable poisons in the fuel pellets filled in the most of upper portion than the upper end of the effective length of the short fuel rod disposed to the outermost circumference of the fuel bundle. (I.N.)

  17. Pressure drop redistribution experimental analysis in axial flow along the bundles

    Fuel elements of PWR type nuclear reactors are composed of rod bundles, arranged in square arrays, held by grid type spacers. The coolant flows axially along the bundle. Although such elements are laterally open, pressure drop experiments are performed in closed type test sections, originating the appearance of subchannels of different geometries. Utilizing a test section of two bundles of 4 x 4 pins and performing experiments with and without separation between the bundles, the flow redistribution factors, the friction, and the grid drag coefficients were determined for the interior subchannels. 03 refs, 06 figs, 02 tabs. (B.C.A.)

  18. Thermal hydraulics of rod bundles: The effect of eccentricity

    Highlights: • Present CFD investigation explores, whole bundle eccentricity for the first time. • Fluid flow and thermal characteristics in various subchannels are analyzed. • Mass flux distribution is particularly analyzed to study eccentricity effect. • Higher eccentricity resulted in a shoot up in rod surface temperature distribution. • Both tangential and radial flow in rod bundles has resulted due to eccentricity. -- Abstract: The effect of eccentricity on the fluid flow and heat transfer through a 19-rod bundle is numerically carried out. When the whole bundle shifts downwards with respect to the outer (pressure) tube, flow redistribution happens. This in turn is responsible for changes in mass flux, pressure and differential flow development in various subchannels. The heat flux imposed on the surface of the fuel rods and the mass flux through the subchannels determines the coolant outlet temperatures. The simulations are performed for a coolant flow Reynolds number of 4 × 105. For an eccentricity value of 0.7, the mass flux in the bottom most subchannel (l) was found to decrease by 10%, while the surface temperature of the fuel rod in the vicinity of this subchannel increased by 250% at the outlet section. Parameters of engineering interest including skin friction coefficient, Nusselt number, etc., have been systematically explored to study the effect of eccentricity on the rod bundle

  19. The Atiyah Bundle and Connections on a Principal Bundle

    Indranil Biswas

    2010-06-01

    Let be a ∞ manifold and a Lie a group. Let $E_G$ be a ∞ principal -bundle over . There is a fiber bundle $\\mathcal{C}(E_G)$ over whose smooth sections correspond to the connections on $E_G$. The pull back of $E_G$ to $\\mathcal{C}(E_G)$ has a tautological connection. We investigate the curvature of this tautological connection.

  20. CFD analysis of flow and heat transfer in Canadian supercritical water reactor bundle

    Highlights: • Flow and heat transfer in SCWR fuel bundle design by AECL is studied using CFD. • Bare-rod bundle geometry is tested at 23.5, 25 and 28 MPa using STAR-CCM+ code. • SST k–ω low-Re model was used to study occurrence of heat transfer deterioration. - Abstract: Within the Gen-IV International Forum, AECL is leading the effort in developing a conceptual design for the Canadian SCWR. AECL proposed a new fuel bundle design with two rings of fuel elements placed between central flow tube and the pressure tube. In line with the scope of the conceptual design, the objective of the present CFD work is to aid in developing a bundle heat transfer correlation for the Canadian SCWR fuel bundle design. This paper presents results from an ongoing effort in determining the conditions favorable for occurrence of HTD in the supercritical bundle flows. In the current investigation, bare-rod bundle geometry was tested for the proposed fuel bundle design at 23.5, 25 and 28 MPa using STAR-CCM+ CFD code. Taking advantage of the design symmetry of the fuel bundle, only 1/32 of the computational domain was simulated. The low-Reynolds number modification of SST k–ω turbulence model along with y+ < 1 was used in the simulations. For lower mass flow simulations, the increase of inlet temperature and operational pressure was found effective in reducing the occurrence of HTD. For higher mass flow simulations, normal heat transfer behaviour was observed except for the lower pressure range (23.5 MPa)

  1. Fuel assembly

    Purpose: To improve the thermal and mechanical safety of fuel rods and structural components by making the local power coefficient of jointed fuel rods greater than that of other fuel rods in a fuel assembly. Constitution: In a fuel assembly comprising a plurality of fuel rods bundled by a spacer and held at the upper and the lower positions with tie plates for insertion into a channel, the degree of enrichment of uranium 235 for uranium dioxide fuel pellets charged in jointed fuel rods is adjusted such that the local power coefficient of the jointed fuel rods is made greater than that of the other fuel rods. In the case if the upper tie plate is moved upwardly by the extension of the jointed fuel rods, other fuel rods axially free from the upper tie plate receives no tension, whereby the safety of the fuel assembly can be improved. (Moriyama, K.)

  2. Left bundle-branch block

    Risum, Niels; Strauss, David; Sogaard, Peter;

    2013-01-01

    The relationship between myocardial electrical activation by electrocardiogram (ECG) and mechanical contraction by echocardiography in left bundle-branch block (LBBB) has never been clearly demonstrated. New strict criteria for LBBB based on a fundamental understanding of physiology have recently...

  3. Bundling ecosystem services in Denmark

    Turner, Katrine Grace; Odgaard, Mette Vestergaard; Bøcher, Peder Klith; Dalgaard, Tommy; Svenning, J.-C.

    2014-01-01

    We made a spatial analysis of 11 ecosystem services at a 10 km × 10 km grid scale covering most of Denmark. Our objective was to describe their spatial distribution and interactions and also to analyze whether they formed specific bundle types on a regional scale in the Danish cultural landscape....... We found clustered distribution patterns of ecosystem services across the country. There was a significant tendency for trade-offs between on the one hand cultural and regulating services and on the other provisioning services, and we also found the potential of regulating and cultural services to...... form synergies. We identified six distinct ecosystem service bundle types, indicating multiple interactions at a landscape level. The bundle types showed specialized areas of agricultural production, high provision of cultural services at the coasts, multifunctional mixed-use bundle types around urban...

  4. Structure of the acrosomal bundle.

    Schmid, Michael F; Sherman, Michael B; Matsudaira, Paul; Chiu, Wah

    2004-09-01

    In the unactivated Limulus sperm, a 60- micro m-long bundle of actin filaments crosslinked by the protein scruin is bent and twisted into a coil around the base of the nucleus. At fertilization, the bundle uncoils and fully extends in five seconds to support a finger of membrane known as the acrosomal process. This biological spring is powered by stored elastic energy and does not require the action of motor proteins or actin polymerization. In a 9.5-A electron cryomicroscopic structure of the extended bundle, we show that twist, tilt and rotation of actin-scruin subunits deviate widely from a 'standard' F-actin filament. This variability in structural organization allows filaments to pack into a highly ordered and rigid bundle in the extended state and suggests a mechanism for storing and releasing energy between coiled and extended states without disassembly. PMID:15343340

  5. Validation of SOCRAT-BN code on rod bundle experiments

    SOCRAT-BN code is developed for the analysis of design and beyond design basis accidents at NPPs with liquid sodium as a coolant. To simulate the behavior of the coolant in the reactor core heat transfer and friction in rod bundle geometry are required to consider. The code SOCRAT-BN uses specialized closing relations to simulate rod bundles. The article describes the validation of the code SOCRAT-BN on experiments with fuel rod imitators in the triangular geometry with a wire-wound. (author)

  6. Fuel assembly

    A fuel assembly is composed of a fuel bundle surrounded by a channel box. The fuel bundle comprises a large number of fuel rods and a water rod secured to upper and lower tie plate by way of a plurality of fuel spacers. Grooves (libretti) are formed in the direction along the flowing direction of coolants to at least one of the surface of the fuel rods, the inner surface of the channel box, the surface of the water rod and spacer constituting components. In this case, the lateral width of the libretto in the flowing direction is determined as the minimum thickness of the bottom layer of a layered flow determined by a coolant flow rate. With such a constitution, abrasion resistance relative to coolants is reduced to reduce the pressure loss of fuel assemblies. (I.N.)

  7. Damage in Fiber Bundle Models

    Kun, Ferenc; Zapperi, Stefano; Herrmann, Hans J.

    1999-01-01

    We introduce a continuous damage fiber bundle model that gives rise to macroscopic plasticity and compare its behavior with that of dry fiber bundles. Several interesting constitutive behaviors are found in this model depending on the value of the damage parameter and on the form of the disorder distribution. In addition, we compare the behavior of global load transfer models with local load transfer models and study in detail the damage evolution before failure. We emphasize the analogies be...

  8. Holomorphic bundles over elliptic manifolds

    In this lecture we shall examine holomorphic bundles over compact elliptically fibered manifolds. We shall examine constructions of such bundles as well as (duality) relations between such bundles and other geometric objects, namely K3-surfaces and del Pezzo surfaces. We shall be dealing throughout with holomorphic principal bundles with structure group GC where G is a compact, simple (usually simply connected) Lie group and GC is the associated complex simple algebraic group. Of course, in the special case G = SU(n) and hence GC = SLn(C), we are considering holomorphic vector bundles with trivial determinant. In the other cases of classical groups, G SO(n) or G = Sympl(2n) we are considering holomorphic vector bundles with trivial determinant equipped with a non-degenerate symmetric, or skew symmetric pairing. In addition to these classical cases there are the finite number of exceptional groups. Amazingly enough, motivated by questions in physics, much interest centres around the group E8 and its subgroups. For these applications it does not suffice to consider only the classical groups. Thus, while often first doing the case of SU(n) or more generally of the classical groups, we shall extend our discussions to the general semi-simple group. Also, we shall spend a good deal of time considering elliptically fibered manifolds of the simplest type, namely, elliptic curves

  9. Analysis of F/M duty cycle and O/M cost for four-bundle shift refuelling scheme in CANDU6 NPP

    A four-bundle shift refuelling method, a refuelling scheme that can reduces local flux peak compared to the current eight-bundle shift refuelling method used in CANDU6 NPP, is analyzed to see how much Fuel Handling System load and management cost increase are required due to the change. The current eight-bundle shift refuelling method requires to refuel eight fuel bundles from a single fuel channel, and to refuel 14 fuel channels in a week on average assuming that the reactor is in a steady state. The four-bundle shift refuelling method increases Fuelling Machine duty cycle and operator load. The study showed that the refuelling scheme change from the eight-to four-bundle shift increases the operation and maintenance cost about 35% from the current figure by conservative estimate and that the Fuel Handling System has enough flexibility to meet the demand of a more frequent refuelling scheme

  10. Does size matter? : disentangling consumers' bundling preferences

    Manoj K. Agarwal; Frambach, Ruud T.; Stremersch, Stefan

    2000-01-01

    Previous marketing literature has focused to a large extent on the effect of bundle characteristics on a consumer’s decision to buy a (fixed) bundle in a non-competitive setting. This study extends this narrow focus in four major ways. First, the authors address bundles that are customizable. Second, they distinguish between a consumer’s decision of whether to bundle (bundle choice) and the decision of how many goods or services to include in a bundle (bundle size). Third, they extend the foc...

  11. Analysis of fuelling sequence and fatigue life for 4-bundle shift refuelling scheme in CANDU6 NPP

    A 4-bundle shift refuelling method of CANDU6 F/H (Fuel Handling) System is analyzed to assess the operational flexibility and capacity of F/H system. The current 8-bundle shift refuelling scheme requires to refuel eight fuel bundles from a single fuel channel, and to refuel 14 fuel channels in a week on average assuming that the reactor is in a steady state. The analysis showed that the 4-bundle shift refuelling method increases F/M (Fuelling Machine) duty cycle and operator load. However, the fuelling method change from the 8- to 4-bundle shift refuelling will not require additional team of operators. A marginal increase in the maintenance cost may be resulted in by the change of fuelling method and the increase of fatigue usage factors requires some components to be replaced during the F/M lifetime

  12. Simulation of bundle test Quench-12 with integral code MELCOR

    The past NRI analyses cover the Quench-01, Quench-03 and Quench-06 with version MELCOR 1.8.5 (including reflood model), and Quench-01 and Quench-11 tests with the latest version MELCOR 1.8.6. The Quench-12 test is specific, because it has different bundle configuration related to the VVER bundle configuration with hexagonal grid of pins and also used E110 cladding material. Specificity of Quench-12 test is also in the used material of fuel rod cladding - E110. The test specificities are a reason for the highest concern, because the VVER reactors are operated in the Czech Republic. The new input model was developed with the taking into account all experience from previous simulations of the Quench bundle tests. The recent version MELCOR 1.8.6 YU2911 was used for the simulation with slightly modified ELHEAT package. Sensitivity studies on input parameters and oxidation kinetics were performed. (author)

  13. The button effect of CANFLEX bundle on the critical heat flux and critical channel power

    Park, Joo Hwan; Jun, Jisu; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G. R.; Bullock, D. E.; Inch, W. [Atomic Energy of Canada Limited, Ontario (Canada)

    1997-12-31

    A CANFLEX (CANdu FLEXible fuelling) 43-element bundle has developed for a CANDU-6 reactor as an alternative of 37-element fuel bundle. The design has two diameter elements (11.5 and 13.5 mm) to reduce maximum element power rating and buttons to enhance the critical heat flux (CHF), compared with the standard 37-element bundle. The freon CHF experiments have performed for two series of CANFLEX bundles with and without buttons with a modelling fluid as refrigerant R-134a and axial uniform heat flux condition. Evaluating the effects of buttons of CANFLEX bundle on CHF and Critical Channel Power (CCP) with the experimental results, it is shown that the buttons enhance CCP as well as CHF. All the CHF`s for both the CANFLEX bundles are occurred at the end of fuel channel with the high dryout quality conditions. The CHF enhancement ratio are increased with increase of dryout quality for all flow conditions and also with increase of mass flux only for high pressure conditions. It indicates that the button is a useful design for CANDU operating condition because most CHF flow conditions for CANDU fuel bundle are ranged to high dryout quality conditions. 5 refs., 11 figs. (Author)

  14. Preliminary assessment of noble gas bundle tagging using a partial krypton backfill

    Current limitations of CANDU reactors to reliably locate defective fuel bundles have sparked interest into new identification techniques. Noble gas tagging, which would involve the addition of specific combinations of Kr and Xe isotopes to the fuel-to-sheath gap during manufacturing, has the potential to offer a means of locating failed-fuel bundles. The released tag with a given isotopic signature could be measured in the primary heat transport system by mass spectrometry. This technique would allow on-power failure location. Moreover, the technique could be of particular interest for demonstration irradiations with new fuel bundle designs. This report outlines preliminary considerations towards a suitable tag isotope choice and discusses the impact on the thermal performance of a fuel element. The detection limit of two mass spectrometer systems was determined through measurements of prepared krypton samples with aqueous concentrations in the range of 10-12 to 10-9 [molKr/molH2O]. (author)

  15. Cohomology of line bundles: Applications

    Blumenhagen, Ralph; Jurke, Benjamin; Rahn, Thorsten; Roschy, Helmut

    2012-01-01

    Massless modes of both heterotic and Type II string compactifications on compact manifolds are determined by vector bundle valued cohomology classes. Various applications of our recent algorithm for the computation of line bundle valued cohomology classes over toric varieties are presented. For the heterotic string, the prime examples are so-called monad constructions on Calabi-Yau manifolds. In the context of Type II orientifolds, one often needs to compute cohomology for line bundles on finite group action coset spaces, necessitating us to generalize our algorithm to this case. Moreover, we exemplify that the different terms in Batyrev's formula and its generalizations can be given a one-to-one cohomological interpretation. Furthermore, we derive a combinatorial closed form expression for two Hodge numbers of a codimension two Calabi-Yau fourfold.

  16. Principal bundles the classical case

    Sontz, Stephen Bruce

    2015-01-01

    This introductory graduate level text provides a relatively quick path to a special topic in classical differential geometry: principal bundles.  While the topic of principal bundles in differential geometry has become classic, even standard, material in the modern graduate mathematics curriculum, the unique approach taken in this text presents the material in a way that is intuitive for both students of mathematics and of physics. The goal of this book is to present important, modern geometric ideas in a form readily accessible to students and researchers in both the physics and mathematics communities, providing each with an understanding and appreciation of the language and ideas of the other.

  17. Thermo-fluid and electrochemical modeling of a multi-bundle IP-SOFC - Technology for second generation hybrid application

    This paper describes an electrochemical model, which studies the performance of multi-bundles integrated-planar solid oxide fuel cell (IP-SOFC strip) fuelled by pure hydrogen. Following a description of the basic geometries and general premises the approaches and simplifications for the calculation of ohmic resistance, gas flow, heat and mass transfers are given. The effect of fuel pressure ratio and the temperature variation are investigated in a parameter study. The model results have been validates well with experimental data obtained from a full-size prototype of IP-SOFC technology for the second generation hybrid application; it is found that the ohmic and cathode activation overpotentials represent a major loss in fuel cell voltage. The results obtained from both strips also demonstrated acceptable performance and good reproducibility bundles to bundles at 900 deg. C. These results indicated weak effect of parameter variation on the bundles to bundles performances.

  18. Thermo-fluid and electrochemical modeling of a multi-bundle IP-SOFC - Technology for second generation hybrid application

    Mounir, H. [L2MCS - Laboratoire de Mecanique des Materiaux et Calcul des Structures de l' ENSET de Rabat, ENSET-Rabat, B.P. 6207, Av. Armee Royale, Badinat Al Irfane Rabat (Morocco); Laboratoire de Thermodynamique et Mecaniques des Materiaux, Faculte des Sciences de Rabat (Morocco); El Gharad, A. [L2MCS - Laboratoire de Mecanique des Materiaux et Calcul des Structures de l' ENSET de Rabat, ENSET-Rabat, B.P. 6207, Av. Armee Royale, Badinat Al Irfane Rabat (Morocco); Belaiche, M. [Laboratoire Magnetisme, Materiaux Magnetiques, Micro-onde et ceramique, Ecole normale superieure, BP 9235, Ocean Rabat 1000 (Morocco); Boukalouch, M. [Laboratoire de Thermodynamique et Mecaniques des Materiaux, Faculte des Sciences de Rabat (Morocco)

    2009-10-15

    This paper describes an electrochemical model, which studies the performance of multi-bundles integrated-planar solid oxide fuel cell (IP-SOFC strip) fuelled by pure hydrogen. Following a description of the basic geometries and general premises the approaches and simplifications for the calculation of ohmic resistance, gas flow, heat and mass transfers are given. The effect of fuel pressure ratio and the temperature variation are investigated in a parameter study. The model results have been validates well with experimental data obtained from a full-size prototype of IP-SOFC technology for the second generation hybrid application; it is found that the ohmic and cathode activation overpotentials represent a major loss in fuel cell voltage. The results obtained from both strips also demonstrated acceptable performance and good reproducibility bundles to bundles at 900 C. These results indicated weak effect of parameter variation on the bundles to bundles performances. (author)

  19. Single and two-phase flow pressure drop for CANFLEX bundle

    Park, Joo Hwan; Jun, Ji Su; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G. R.; Bullock, D. E. [Atomic Energy of Canada Limited, Ontario (Canada)

    1998-12-31

    Friction factor and two-phase flow frictional multiplier for a CANFLEX bundle are newly developed and presented in this paper. CANFLEX as a 43-element fuel bundle has been developed jointly by AECL/KAERI to provide greater operational flexibility for CANDU reactor operators and designers. Friction factor and two-phase flow frictional multiplier have been developed by using the experimental data of pressure drops obtained from two series of Freon-134a (R-134a) CHF tests with a string of simulated CANFLEX bundles in a single phase and a two-phase flow conditions. The friction factor for a CANFLEX bundle is found to be about 20% higher than that of Blasius for a smooth circular pipe. The pressure drop predicted by using the new correlations of friction factor and two-phase frictional multiplier are well agreed with the experimental pressure drop data of CANFLEX bundle within {+-} 5% error. 11 refs., 5 figs. (Author)

  20. Exploring Bundling Theory with Geometry

    Eckalbar, John C.

    2006-01-01

    The author shows how instructors might successfully introduce students in principles and intermediate microeconomic theory classes to the topic of bundling (i.e., the selling of two or more goods as a package, rather than separately). It is surprising how much students can learn using only the tools of high school geometry. To be specific, one can…

  1. Bundled Discounts and EC Judicial Review

    Christian Roques

    2008-01-01

    The Community Courts' case law is rich with cases relating to tying or bundling practices in their classical economic form. However, the same cannot be said for the second acceptance of bundled discounts.

  2. Heat transfer in rod bundles with severe clad deformations

    The content of the paper is focused on heat transfer conditions during the reflood phase of a LOCA in slightly to severely deformed PWR fuel rod bundle geometries. The status of analytical and, especially, of experimental work is described as far as it is possible within this frame. Emphasis is placed on the presentation of the results of ''Flooding Experiments with Blocked Arrays'' (FEBA), a program performed at the Kernforschungszentrum Karlsruhe in the frame of the Project Nuclear Safety (PNS). (orig./WL)

  3. Failure properties of fiber bundle models

    Pradhan, Srutarshi; Chakrabarti, Bikas K.

    2003-01-01

    We study the failure properties of fiber bundles when continuous rupture goes on due to the application of external load on the bundles. We take the two extreme models: equal load sharing model (democratic fiber bundles) and local load sharing model. The strength of the fibers are assumed to be distributed randomly within a finite interval. The democratic fiber bundles show a solvable phase transition at a critical stress (load per fiber). The dynamic critical behavior is obtained analyticall...

  4. Bundling Information Goods: Pricing, Profits, and Efficiency

    Yannis Bakos; Erik Brynjolfsson

    1999-01-01

    We study the strategy of bundling a large number of information goods, such as those increasingly available on the Internet, and selling them for a fixed price. We analyze the optimal bundling strategies for a multiproduct monopolist, and we find that bundling very large numbers of unrelated information goods can be surprisingly profitable. The reason is that the law of large numbers makes it much easier to predict consumers' valuations for a bundle of goods than their valuations for the indi...

  5. Fuel assembly for a boiling water reactor

    A fuel assembly for a boiling water reactor comprises a plurality of fuel rods which constitute four partial bundles and are surrounded by a fuel channel system comprising one partial tube for each partial bundle. Each of the four partial bundles rests on a bottom tie plate and is positioned with respect to the others by means of a common top tie plate which is provided with a lifting loop which is sufficiently strong to be able to lift the four partial bundles simultaneously, a major part of the lifting force being transmitted to said bottom tie plates via a plurality of supporting fuel rods

  6. An experimental assessment of cooling of a 54-rod bundle by in-bundle injection

    Highlights: ► Rewetting of a 54-rod bundle assembled with a central coolant tube is investigated. ► The coolant tube injects the coolant radially outwards at different axial levels. ► Above a minimum flow rate, coolant quenches all the rods throughout their length. ► Rapid cooling of rods occurs up to around 100 °C of the rod surface temperature. ► Counter current flow of steam–water gets generated which affects cooling adversely. - Abstract: The performance of an in-bundle coolant injection system for the quenching of dry heated rods has been experimentally investigated. The rod bundle contains 54 fuel rods of 11.2 mm diameter, 3700 mm long, arranged in three concentric rings with a central coolant supply tube. The coolant tube supplies the coolant in the form of jets through a series of circumferential holes at different axial levels inside the rod bundle. Visualization during cold state injection tests ensures that the liquid spray can reach different levels of all the rods above a certain flow rate of water through the coolant tube. Extensive cooling experiments were done to assess the suitability of the proposed scheme of in-bundle coolant injection. Time–temperature curves have been derived from rods at different locations, from different heights of the rods, over a range of coolant flow rate as well as for different rod temperatures. The effect of the presence of the spacers on local cooling has also been investigated. The cooling curves follow a general trend of a rapid temperature drop up to almost 100 °C of the rod surface temperature irrespective of the operating parameters and the location of the rod. Thereafter, the temperature falls slowly reaching the coolant temperature almost asymptotically. Moreover, the second phase of cooling is often marked by temperature fluctuations of random nature. It was also observed that a large volume of steam generates during cooling and comes out through the top of the test section expelling a

  7. Quantum principal bundles and corresponding gauge theories

    Durdevic, M

    1995-01-01

    A generalization of classical gauge theory is presented, in the framework of a noncommutative-geometric formalism of quantum principal bundles over smooth manifolds. Quantum counterparts of classical gauge bundles, and classical gauge transformations, are introduced and investigated. A natural differential calculus on quantum gauge bundles is constructed and analyzed. Kinematical and dynamical properties of corresponding gauge theories are discussed.

  8. Strategic and welfare implications of bundling

    Martin, Stephen

    1999-01-01

    A standard oligopoly model of bundling shows that bundling by a firm with a monopoly over one product has a strategic effect because it changes the substitution relationships between the goods among which consumers choose. Bundling in appropriate proportions is privately profitable, reduces rivals...

  9. On Volumes of Arithmetic Line Bundles

    Yuan, Xinyi

    2008-01-01

    We show an arithmetic generalization of the recent work of Lazarsfeld-Mustata which uses Okounkov bodies to study linear series of line bundles. As applications, we derive a log-concavity inequality on volumes of arithmetic line bundles and an arithmetic Fujita approximation theorem for big line bundles.

  10. NUCLEAR REACTOR FUEL ELEMENT

    Wheelock, C.W.; Baumeister, E.B.

    1961-09-01

    A reactor fuel element utilizing fissionable fuel materials in plate form is described. This fuel element consists of bundles of fuel-bearing plates. The bundles are stacked inside of a tube which forms the shell of the fuel element. The plates each have longitudinal fins running parallel to the direction of coolant flow, and interspersed among and parallel to the fins are ribs which position the plates relative to each other and to the fuel element shell. The plate bundles are held together by thin bands or wires. The ex tended surface increases the heat transfer capabilities of a fuel element by a factor of 3 or more over those of a simple flat plate.

  11. Experimental analysis of redistribution of the transversal crossflow in rod bundles

    Fuel elements for PWR type nuclear reactors consist of rod bundles, in a square array and are held by grids. The coolant flows, mainly, axially along the rods. The inlet flow bad distribution can yield a strong crossflow. The present work consists in the experimental analysis of the transversal crossflow between 2 bundles with 4x4 rods each, with and without the presence of spacer-type grids, for several inlet flow conditions. It was observed that the crossflow is strongly dependent of the static pressure difference between the bundles and that the presence of grids induces a rapid homogenization of the flow. (C.M.)

  12. Reflood Phenomena in a 5 x 5 Ballooned Rod Bundle

    Kim, Byoung Jae; Kim, Jong Rok; Kim, Kihwan; Moon, S. K. [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Various experimental programs were carried out for the coolability of an assembly containing a partial blockage in a group of ballooned fuel rods under LOCA conditions. A review on these experimental programs is well documented in. One key distinguished feature of KAERI research activities is the consideration of local power increase owing to fuel relocation, whereas the past experimental program did not consider the effect of fuel relocation. The purpose of this study is to investigate the reflood phenomena in the partial blocked 5 x 5 rod bundle. A series of the forced reflood tests were performed with/without consideration of local power increase by fuel relocation. The experimental data were evaluated with numerical predictions using MARS code. The flow blockage alone has little effect on the peak wall temperature. However, the local power increase by fuel relocation affects considerably the peak wall temperature and the time period during which high wall temperatures continue.

  13. Reflood Phenomena in a 5 x 5 Ballooned Rod Bundle

    Various experimental programs were carried out for the coolability of an assembly containing a partial blockage in a group of ballooned fuel rods under LOCA conditions. A review on these experimental programs is well documented in. One key distinguished feature of KAERI research activities is the consideration of local power increase owing to fuel relocation, whereas the past experimental program did not consider the effect of fuel relocation. The purpose of this study is to investigate the reflood phenomena in the partial blocked 5 x 5 rod bundle. A series of the forced reflood tests were performed with/without consideration of local power increase by fuel relocation. The experimental data were evaluated with numerical predictions using MARS code. The flow blockage alone has little effect on the peak wall temperature. However, the local power increase by fuel relocation affects considerably the peak wall temperature and the time period during which high wall temperatures continue

  14. Reversible BWR fuel assembly and method of using same

    A nuclear fuel assembly is described comprising: (a) a flow channel; (b) a lower nozzle assembly structurally attached to the flow channel to form therewith an external envelope; (c) an invertible fuel bundle adapted to be inserted into the envelope, the fuel bundle comprising elongated fuel rods held in a spaced lateral array between top and bottom tie plates. Each of the top and bottom tie plates is substantially identical and has means for supporting the fuel bundle within the envelope in either of two mutually inverted vertical orientations whereby the orientation of the fuel bundle in a flow channel may be reversed during burn-up operation

  15. The management status of the spent fuel in HANARO(1995-2009)

    In HANARO, the spent fuels are stored in the spent fuel storage pool of the reactor hall. The capacity of the spent fuel storage pool was designed to store 600 bundles for 36 rods fuel, 432 bundles for 18 rods fuel, 315 rods for TRIGA reactor fuel and the fuels loaded in the reactor core. The spent fuel storage pool can store spent fuels discharged from the reactor core for 20 years normal operation. As for July 2009, the spent fuel 337 bundles are stored in the spent fuel storage pool. There are 217 bundles of 36 rods fuel and 120 bundles of 18 rods fuel. In this report, the information of the spent fuel about the loading date in the reactor core, discharged date, burnup, invisible inspection results and loading position in the spent fuel storage pool are described

  16. Simplicial principal bundles in parametrized spaces

    Roberts, David M

    2012-01-01

    In this paper, motivated by recent interest in higher gauge theory, we prove that the fiberwise geometric realization functor takes a certain class of simplicial principal bundles in a suitable category of spaces over a fixed space $B$ to fiberwise principal bundles. As an application we show that the fiberwise geometric realization of the universal simplicial principal bundle for a simplicial group $G$ in the category of spaces over $B$ gives rise to a fiberwise principal bundle with structure group $|G|$. An application to classifying theory for fiberwise principal bundles is described.

  17. Multipath packet switch using packet bundling

    Berger, Michael Stubert

    The basic concept of packet bundling is to group smaller packets into larger packets based on, e.g., quality of service or destination within the packet switch. This paper presents novel applications of bundling in packet switching. The larger packets created by bundling are utilized to extend...... switching capacity by use of parallel switch planes. During the bundling operation, packets will experience a delay that depends on the actual implementation of the bundling and scheduling scheme. Analytical results for delay bounds and buffer size requirements are presented for a specific scheduling...

  18. Concept and objectives of diagnostic rod bundle development for WWER-440 and account of carried out steps

    The advantages of an in-reactor monitoring system as compared with external control and measuring devices are discussed. A short historic retrospective of the in-reactor monitorino. system development is given with the stress on unresolved problems. Realization of such a system on the basis of a diagnostic fuel rod bundle with controlled coolant flow rats is considered. The scientific objectives of the diagnostic rod bundle development proo.ramme are identified. For DK1 and DK2 diagnostic rod bundles specification lists of measured parameters and corresponding sensors are presented. Problems of advanced design diagnostic fuel rod bundles (DK3 family) development are discussed. DK1 - 2 bundle is described in detail as well as the data acquisition and processing system and the results of experiments on Rheinsberg NPP

  19. Model of turbine blades bundles

    Půst, Ladislav; Pešek, Luděk

    Prague : Institute of Thermomechanics, Academy of Sciences of the Czech Republic, v. v. i., 2013 - (Zolotarev, I.), s. 467-477 ISBN 978-80-87012-47-5. ISSN 1805-8256. [Engineering Mechanics 2013 /19./. Svratka (CZ), 13.05.2013-16.05.2013] R&D Projects: GA ČR GA101/09/1166 Institutional support: RVO:61388998 Keywords : free and forced vibrations * eigenmodes * mathematical model * bundle of blades Subject RIV: BI - Acoustics

  20. Model of turbine blades bundles

    Půst, Ladislav; Pešek, Luděk

    Praha : Insitute of Thermomechanics ASCR, v. v. i., 2013 - (Zolotarev, I.). s. 125-126 ISBN 978-80-87012-46-8. ISSN 1805-8248. [Engineering Mechanics 2013 /19./. 13.05.2013-16.05.2013, Svratka] R&D Projects: GA ČR GA101/09/1166 Institutional support: RVO:61388998 Keywords : free and forced vibrations * eigenmodes * bundle of blades Subject RIV: BI - Acoustics

  1. Competitive nonlinear pricing and bundling

    Armstrong, Mark; Vickers, John

    2006-01-01

    We examine the impact of multiproduct nonlinear pricing on profit, consumer surplus and welfare in a duopoly. When consumers buy all their products from one firm (the one-stop shopping model), nonlinear pricing leads to higher profit and welfare, but often lower consumer surplus, than linear pricing. By contrast, in a unit-demand model where consumers may buy one product from one firm and another product from another firm, bundling generally acts to reduce profit and welfare and to boost cons...

  2. Quantum bundles and their symmetries

    Wave functions in the domain of observables such as the Hamiltonian are not always smooth functions on the classical configuration space Q. Rather, they are often best regarded as functions on a G bundle EG over Q or as sections of an associated bundle. If H is a classical group which acts on Q, its quantum version HG, which acts on EG, is not always H, but an extension of H by G. A powerful and physically transparent construction of EG and HG, where G = U(1) and H1(Q,Z) = 0, has been developed using the path space P. (P consists of paths on Q from a fixed point). In this paper the authors show how to construct EG and HG when G is U(1) or U(1) x π1(Q) and there is no restriction on H1(Q,Z). The method is illustrated with concrete examples, such as a system of charges and monopoles. The method is illustrated with concrete examples, such as a system of charges and monopoles. The authors argue also that P is a sort of superbundle from which a large variety of bundles can be obtained by imposing suitable equivalence relations

  3. Photonic bandgap fiber bundle spectrometer

    Hang, Qu; Syed, Imran; Guo, Ning; Skorobogatiy, Maksim

    2010-01-01

    We experimentally demonstrate an all-fiber spectrometer consisting of a photonic bandgap fiber bundle and a black and white CCD camera. Photonic crystal fibers used in this work are the large solid core all-plastic Bragg fibers designed for operation in the visible spectral range and featuring bandgaps of 60nm - 180nm-wide. 100 Bragg fibers were chosen to have complimentary and partially overlapping bandgaps covering a 400nm-840nm spectral range. The fiber bundle used in our work is equivalent in its function to a set of 100 optical filters densely packed in the area of ~1cm2. Black and white CCD camera is then used to capture spectrally "binned" image of the incoming light at the output facet of a fiber bundle. To reconstruct the test spectrum from a single CCD image we developed an algorithm based on pseudo-inversion of the spectrometer transmission matrix. We then study resolution limit of this spectroscopic system by testing its performance using spectrally narrow test peaks (FWHM 5nm-25nm) centered at va...

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

    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

  5. CFD analysis of flow field in a triangular rod bundle

    The flow field was investigated in subchannels of VVER-440 pressurized water cooled reactors' fuel assemblies (triangular lattice, P/D = 1.35). Impacts of the mesh resolution and turbulence model were studied in order to obtain guidelines for CFD calculations of VVER-440 rod bundles. Results were compared to measurement data published by Trupp and Azad in 1975. The study pointed out that RANS method with BSL Reynolds stress model using a sufficient fine grid can provide an accurate prediction for the turbulence quantities in this lattice. Applying the experiences of the sensitivity study thermal hydraulic processes were investigated in VVER-440 rod bundle sections. Based on the examinations the spacer grids have important effects on the cross flows, axial velocity and outlet temperature distribution of subchannels therefore they have to be modeled satisfactorily in CFD calculations.

  6. Subchannel Analysis for enhancing the fuel performance in CANDU reactor

    The effect of the fuel rod geometry in a fuel bundle using the subchannel code ASSERT has been evaluated to design the fuel bundle having the advanced fuel performance. Based on the configuration of standard 37-element fuel bundle, the element diameter of fuel rods in each ring has been changed while that of fuel rods in other rings is kept as the original size. The dryout power of each element in a fuel bundle has been obtained for the modified fuel bundle and compared with that of a standard fuel bundle. From the calculated mixture enthalpy and void fraction of each subchannel, it was found that the modification of element diameter largely affects to the thermal characteristics of the subchannel on the upper region of a modified element by the buoyancy drift effect. The optimized geometry in a fuel bundle has been suggested from the consideration of the change of void reactivity as well as the dryout power of a bundle. The dependency of the transverse interchange model on the present results has been checked by examining the dryout power of a bundle for the different mixing coefficient and buoyancy drift model

  7. Fabrication of a CANFLEX-RU designed bundle for power ramp irradiation test in NRU

    The BDL-443 CANFLEX-RU bundle AKW was fabricated at Korea Atomic Energy Research Institute (KAERI) for power ramp irradiation testing in NRU reactor. The bundle was fabricated with IDR and ADU fuel pellets in adjacent elements and contains fuel pellets enriched to 1.65 wt% 235U in the outer and intermediate rings and also contains pellets enriched to 2.00 wt% 235U in the inner ring. This bundle does not have a center element to allow for insertion on a hanger bar. KAERI produced the IDR pellets with the IDR-source UO2 powder supplied by BNFL. ADU pellets were fabricated and supplied by AECL. Bundle kits (Zircaloy-4 end plates, end plugs, and sheaths with brazed appendages) manufactured at KAERI earlier in 1996 were used for the fabrication of the bundle. The CANFLEX bundle was fabricated successfully at KAERI according to the QA provisions specified in references and as per relevant KAERI drawings and technical specification. This report covers the fabrication activities performed at KAERI. Fabrication processes performed at AECL will be documented in a separate report

  8. Global analysis of bundle behavior in pressurized water reactor specific CORA experiments

    At Kernforschungszentrum Karlsruhe, out-of-pile bundle experiments are performed in the CORA facility to investigate the behavior of light water reactor fuel elements during severe fuel damage accidents. To analyze the phenomena observed during the tests, such as claddin failure, oxidation, and deformation, as well as their influence on the post test bundle state, four pressurized water reactor specific tests are selected: CORA-2, CORA-3, CORA-5, and CORA-12. From each of these tests, a detailed global analysis using all the measured temperatures, pressures, and fluid compositions as well as videoscope information has been performed. To describe the post test bundle state quantitatively, axial profiles of the bundle cross-section area, the damage state of the rods, the average cladding oxidation, and the damage to the pellets are measured. The effects of CORA-specific components on the bundle melt progression and the measured axial profiles are identified and assessed. Most of the observations during the tests as well as the post test bundle state can be explained by the established common sequence of phenomena. For a better understanding of the melt progression, some physical phenomena, such as the energy release associated with the double-sided oxidation of the cladding, the melt release, or the melt relocation, must be analyzed in detail

  9. Quadratic bundle and nonlinear equations

    The paper is aimed at giving an exhaustive description of the nonlinear evolution equations (NLEE), connected with the quadratic bundle (the spectral parameter lambda, which enters quadratically into the equations) and at describing Hamiltonian structure of these equations. The equations are solved through the inverse scattering method (ISM). The basic formulae for the scattering problem are given. The spectral expansion of the integrodifferential operator is used so that its eigenfunctions are the squared solutions of the equation. By using the notions of Hamiltonian structure hierarchy and gauge transformations it is shown how to single out physically interesting NLEE

  10. Damping Properties of the Hair Bundle

    Baumgart, Johannes; Kozlov, Andrei S.; Risler, Thomas; Hudspeth, A. James

    2015-01-01

    The viscous liquid surrounding a hair bundle dissipates energy and dampens oscillations, which poses a fundamental physical challenge to the high sensitivity and sharp frequency selectivity of hearing. To identify the mechanical forces at play, we constructed a detailed finite-element model of the hair bundle. Based on data from the hair bundle of the bullfrog's sacculus, this model treats the interaction of stereocilia both with the surrounding liquid and with the liquid in the narrow gaps b...

  11. Tying, Bundling, and Loyalty/Requirement Rebates

    Nicholas Economides

    2011-01-01

    I discuss the impact of tying, bundling, and loyalty/requirement rebates on consumer surplus in the affected markets. I show that the Chicago School Theory of a single monopoly surplus that justifies tying, bundling, and loyalty/requirement rebates on the basis of efficiency typically fails. Thus, tying, bundling, and loyalty/requirement rebates can be used to extract consumer surplus and enhance profit of firms with market power. I discuss the various setups when this occurs.

  12. Bundling and Competition on the Internet

    Yannis Bakos; Erik Brynjolfsson

    2000-01-01

    The Internet has signi.cantly reduced the marginal cost of producing and distributing digital information goods. It also coincides with the emergence of new competitive strategies such as large-scale bundling. In this paper, we show that bundling can create “economies of aggregation” for information goods if their marginal costs are very low, even in the absence of network externalities or economies of scale or scope. We extend the Bakos-Brynjolfsson bundling model (1999) to settings with sev...

  13. Bundling and joint marketing by rival firms

    Jeitschko, Thomas D.; Jung, Yeonjei; Kim, Jaesoo

    2014-01-01

    We study joint marketing arrangements by competing firms who engage in price discrimination between consumers who patronize only one firm (single purchasing) and those who purchase from both competitors (bundle purchasers). Two types of joint marketing are considered. Firms either commit to a component-price that applies to bundle-purchasers and then firms set stand-alone prices for single purchasers; or firms commit to a rebate off their stand alone price that will be applied to bundle-purch...

  14. DP-THOT - a calculational tool for bundle-specific decay power based on actual irradiation history

    A tool has been created for calculating the decay power of an individual fuel bundle to take account of its actual irradiation history, as tracked by the fuel management code SORO. The DP-THOT tool was developed in two phases: first as a standalone executable code for decay power calculation, which could accept as input an entirely arbitrary irradiation history; then as a module integrated with SORO auxiliary codes, which directly accesses SORO history files to retrieve the operating power history of the bundle since it first entered the core. The methodology implemented in the standalone code is based on the ANSI/ANS-5.1-1994 formulation, which has been specifically adapted for calculating decay power in irradiated CANDU reactor fuel, by making use of fuel type specific parameters derived from WIMS lattice cell simulations for both 37 element and 28 element CANDU fuel bundle types. The approach also yields estimates of uncertainty in the calculated decay power quantities, based on the evaluated error in the decay heat correlations built-in for each fissile isotope, in combination with the estimated uncertainty in user-supplied inputs. The method was first implemented in the form of a spreadsheet, and following successful testing against decay powers estimated using the code ORIGEN-S, the algorithm was coded in FORTRAN to create an executable program. The resulting standalone code, DP-THOT, accepts an arbitrary irradiation history and provides the calculated decay power and estimated uncertainty over any user-specified range of cooling times, for either 37 element or 28 element fuel bundles. The overall objective was to produce an integrated tool which could be used to find the decay power associated with any identified fuel bundle or channel in the core, taking into account the actual operating history of the bundles involved. The benefit is that the tool would allow a more realistic calculation of bundle and channel decay powers for outage heat sink planning

  15. Statistical Constitutive Equation of Aramid Fiber Bundles

    熊杰; 顾伯洪; 王善元

    2003-01-01

    Tensile impact tests of aramid (Twaron) fiber bundles were carried om under high strain rates with a wide range of 0. 01/s~1000/s by using MTS and bar-bar tensile impact apparatus. Based on the statistical constitutive model of fiber bundles, statistical constitutive equations of aramid fiber bundles are derived from statistical analysis of test data at different strain rates. Comparison between the theoretical predictions and experimental data indicates statistical constitutive equations fit well with the experimental data, and statistical constitutive equations of fiber bundles at different strain rates are valid.

  16. Extension of holomorphic bundles to the disc (and Serre's Problem on Stein bundles)

    Rosay, Jean-Pierre

    2006-01-01

    We show how to extend some holomorphic bundles with fifer C^2 and base an open set in C, to bundles on the Riemann Sphere, by an extremely simple technique. In particular, it applies to examples of non-Stein bundles constructed by Skoda and Demailly. It gives an example on C, with polynomial transition automorphisms.

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

    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.

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

    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

  19. An analytical model for the prediction of fluid-elastic forces in a rod bundle subjected to axial flow: theory, experimental validation and application to PWR fuel assemblies; Calcul des forces fluidelastiques dans les faisceaux de tubes sous ecoulement axial: theorie, validation, application aux assemblages combustibles des REP

    Beaud, F. [Electricite de France (EDF), 78 - Chatou (France)

    1997-12-31

    A model predicting the fluid-elastic forces in a bundle of circular cylinders subjected to axial flow is presented in this paper. Whereas previously published models were limited to circular flow channel, the present one allows to take a rectangular flow external boundary into account. For that purpose, an original approach is derived from the standard method of images. This model will eventually be used to predict the fluid-structure coupling between the flow of primary coolant and a fuel assemblies in PWR nuclear reactors. It is indeed of major importance since the flow is shown to induce quite high damping and could therefore mitigate the incidence of an external load like a seismic excitation on the dynamics of the assemblies. The proposed model is validated on two cases from the literature but still needs further comparisons with the experiments being currently carried out on the EDF set-up. The flow has been shown to induce an approximate 12% damping on a PWR fuel assembly, at nominal reactor conditions. The possible grid effect on the fluid-structure coupling has been neglected so far but will soon be investigated at EDF. (author). 16 refs.

  20. Experimental studies on in-bundle ECCS injection for Advanced Heavy Water Reactor

    The Advanced Heavy Water Reactor (AHWR) being designed at BARC is an innovative reactor with Thorium utilization as its major objective. It has many advanced passive safety features. One such feature is passive injection of emergency coolant after postulated Loss of Coolant Accident (LOCA). A novel feature of this injection scheme is that the injection does not take place in the header/plenum as in other reactors, but directly in to the bundle. For this purpose, the fuel cluster incorporates a central water rod which communicates with the ECCS header. The water rod extends along full length of the fuel cluster. In event of LOCA in the Main Heat Transport (MHT) system, ECC water flows from the accumulator to the water rod through ECCS header. The water flows into the bundle through holes in the water rod. The AHWR fuel cluster has fuel pins arranged in three concentric rings (of 12, 18 and 24 pins) around the central rod. While it is ensured that water does reach the fuel cluster, whether it reaches the outer ring of pins is needs investigation as the pins are closely spaced (1-3 mm gap between adjacent rods). The objective of the present experiments is to determine under what conditions (ECC flow and decay heat), the ECC water is able to rewet and cool all the fuel pins. The experiments have been done in a short, instrumented fuel bundle simulating the geometry of the AHWR fuel cluster

  1. Severe fuel damage scoping test postirradiation examination results

    The fuel bundle from the Severe Fuel Damage Scoping Test, conducted in the Power Burst Facility as part of the international Severe Fuel Damage Research Program, was examined posttest. This paper presents the results of the nondestructive portion of the examination, including gross gamma scanning, neutron radiography, and tomographic reconstruction of cross sections through the bundle using the neutron radiographs

  2. Principal Bundles on the Projective Line

    V B Mehta; S Subramanian

    2002-08-01

    We classify principal -bundles on the projective line over an arbitrary field of characteristic ≠ 2 or 3, where is a reductive group. If such a bundle is trivial at a -rational point, then the structure group can be reduced to a maximal torus.

  3. The Verlinde formula for Higgs bundles

    Andersen, Jørgen Ellegaard; Pei, Du

    2016-01-01

    We propose and prove the Verlinde formula for the quantization of the Higgs bundle moduli spaces and stacks for any simple and simply-connected group. This generalizes the equivariant Verlinde formula for the case of $SU(n)$ proposed previously by the second and third author. We further establish a Verlinde formula for the quantization of parabolic Higgs bundle moduli spaces and stacks.

  4. k-Gerbes, Line Bundles and Anomalies

    Ekstrand, C

    2000-01-01

    We use sets of trivial line bundles for the realization of gerbes. For1-gerbes the structure arises naturally for the Weyl fermion vacuum bundle at afixed time. The Schwinger term is an obstruction in the triviality of a1-gerbe.

  5. k-Gerbes, Line Bundles and Anomalies

    We use sets of trivial line bundles for the realization of gerbes. For 1-gerbes the structure arises naturally for the Weyl fermion vacuum bundle at a fixed time. The Schwinger term is an obstruction in the triviality of a 1-gerbe. (author)

  6. k-Gerbes, Line Bundles and Anomalies

    Ekstrand, C.

    2000-01-01

    We use sets of trivial line bundles for the realization of gerbes. For 1-gerbes the structure arises naturally for the Weyl fermion vacuum bundle at a fixed time. The Schwinger term is an obstruction in the triviality of a 1-gerbe.

  7. Heights for line bundles on arithmetic surfaces

    Jahnel, Joerg

    1995-01-01

    For line bundles on arithmetic varieties we construct height functions using arithmetic intersection theory. In the case of an arithmetic surface, generically of genus g, for line bundles of degree g equivalence is shown to the height on the Jacobian defined by the Theta divisor.

  8. Damping Properties of the Hair Bundle

    Baumgart, Johannes; Kozlov, Andrei S.; Risler, Thomas; Hudspeth, A. J.

    2011-11-01

    The viscous liquid surrounding a hair bundle dissipates energy and dampens oscillations, which poses a fundamental physical challenge to the high sensitivity and sharp frequency selectivity of hearing. To identify the mechanical forces at play, we constructed a detailed finite-element model of the hair bundle. Based on data from the hair bundle of the bullfrog's sacculus, this model treats the interaction of stereocilia both with the surrounding liquid and with the liquid in the narrow gaps between the individual stereocilia. The investigation revealed that grouping stereocilia in a bundle dramatically reduces the total drag. During hair-bundle deflections, the tip links potentially induce drag by causing small but very dissipative relative motions between stereocilia; this effect is offset by the horizontal top connectors that restrain such relative movements at low frequencies. For higher frequencies the coupling liquid is sufficient to assure that the hair bundle moves as a unit with a low total drag. This work reveals the mechanical characteristics originating from hair-bundle morphology and shows quantitatively how a hair bundle is adapted for sensitive mechanotransduction.

  9. Fock modules and noncommutative line bundles

    Landi, Giovanni

    2016-09-01

    To a line bundle over a noncommutative space there is naturally associated a Fock module. The algebra of corresponding creation and annihilation operators is the total space algebra of a principal U(1) -bundle over the noncommutative space. We describe the general construction and illustrate it with examples.

  10. Local thermal-hydraulic behaviour in tight 7-rod bundles

    Cheng, X. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Dongchuan Road 800, 200240 Shanghai (China); Institute for Nuclear and Energy Technologies, Research Centre Karlsruhe, Postfach 3640, 76021 Karlsruhe (Germany)], E-mail: chengxu@sjtu.edu.cn; Yu, Y.Q. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Dongchuan Road 800, 200240 Shanghai (China)

    2009-10-15

    Advanced water-cooled reactor concepts with tight lattices have been proposed worldwide to improve the fuel utilization and the economic competitiveness. In the present work, experimental investigations were performed on thermal-hydraulic behaviour in tight hexagonal 7-rod bundles under both single-phase and two-phase conditions. Freon-12 was used as working fluid due to its convenient operating parameters. Tests were carried out under both single-phase and two-phase flow conditions. Rod surface temperatures are measured at a fixed axial elevation and in various circumferential positions. Test data with different radial power distributions are analyzed. Measured surface temperatures of unheated rods are used for the assessment of and comparison with numerical codes. In addition, numerical simulation using sub-channel analysis code MATRA and the computational fluid dynamics (CFD) code ANSYS-10 is carried out to understand the experimental data and to assess the validity of these codes in the prediction of flow and heat transfer behaviour in tight rod bundle geometries. Numerical results are compared with experimental data. A good agreement between the measured temperatures on the unheated rod surface and the CFD calculation is obtained. Both sub-channel analysis and CFD calculation indicates that the turbulent mixing in the tight rod bundle is significantly stronger than that computed with a well established correlation.

  11. Dirac structures and Dixmier-Douady bundles

    Alekseev, A

    2009-01-01

    A Dirac structure on a vector bundle V is a maximal isotropic subbundle E of the direct sum of V with its dual. We show how to associate to any Dirac structure a Dixmier-Douady bundle A, that is, a Z/2Z-graded bundle of C*-algebras with typical fiber the compact operators on a Hilbert space. The construction has good functorial properties, relative to Morita morphisms of Dixmier-Douady bundles. As applications, we show that the `spin' Dixmier-Douady bundle over a compact, connected Lie group (as constructed by Atiyah-Segal) is multiplicative, and we obtain a canonical `twisted Spin-c-structure' on spaces with group valued moment maps.

  12. Line bundle embeddings for heterotic theories

    Nibbelin, Stefan Groot; Ruehle, Fabian

    2016-04-01

    In heterotic string theories consistency requires the introduction of a non-trivial vector bundle. This bundle breaks the original ten-dimensional gauge groups E8 × E8 or SO(32) for the supersymmetric heterotic string theories and SO(16) × SO(16) for the non-supersymmetric tachyon-free theory to smaller subgroups. A vast number of MSSM-like models have been constructed up to now, most of which describe the vector bundle as a sum of line bundles. However, there are several different ways of describing these line bundles and their embedding in the ten-dimensional gauge group. We recall and extend these different descriptions and explain how they can be translated into each other.

  13. Line bundle embeddings for heterotic theories

    Nibbelink, Stefan Groot

    2016-01-01

    In heterotic theories consistency requires the introduction of a non-trivial vector bundle. This bundle breaks the original ten-dimensional gauge groups E_8 x E_8 or SO(32) for the supersymmetric heterotic theories and SO(16) x SO(16) for the non-supersymmetric tachyon-free theory to smaller subgroups. A vast number of MSSM-like models have been constructed up to now, most of which describe the vector bundle as a sum of line bundles. However, there are several different ways of describing these line bundles and their embedding in the ten-dimensional gauge group. We recall and extend these different descriptions and explain how they can be translated into each other.

  14. Parameter design and optimization of tight-lattice rod bundles

    Thin rod bundles with tight lattice are arranged according to the equilateral triangle grid, as the proportion of fuel is large, and the power density of core is high. Based on the analysis of the performance of core, the ABV-6M reactor is taken as the example, and two objective functions, power density and flow rate of coolant are proposed for optimization calculation. Diameter and pitch of rod are optimized by using GA method respectively. The results, which are considered to be safety in security checking, show that tight lattice is effective for improving the power density and other performances of the reactor core. (author)

  15. Canonical singular hermitian metrics on relative logcanonical bundles

    Tsuji, Hajime

    2010-01-01

    This supersedes 0704.0566. We prove the invariance of logarithmic plurigenera for a projective family of KLT pairs and the adjoint line bundle of KLT line bundles. The proof uses the canonical singular hermitian metrics on relative logcanonical bundles.

  16. On Harder–Narasimhan Reductions for Higgs Principal Bundles

    Arijit Dey; R Parthasarathi

    2005-05-01

    The existence and uniqueness of – reduction for the Higgs principal bundles over nonsingular projective variety is shown. We also extend the notion of – reduction for (, )-bundles and ramified -bundles over a smooth curve.

  17. Gauge symmetries and fibre bundles

    The matter is organized as follows. After a brief introduction to the concept of gauge invariance and its relationship to determinism, we introduce in chapters 3 and 4 the notion of fibre bundles in the context of a discussion on spinning point particles and Dirac monopoles. Chapter 3 deals with a non relativistic treatment of the spinning particle. The non trivial extension to relativistic spinning particles is dealt with in Chapter 5. The free particle system as well as interactions with external electromagnetic and gravitational fields are discussed in detail. In chapter 5 we also elaborate on a remarkable relationship between the charge-monopole system and the system of a massless particle with spin. The classical description of Yang-Mills particles with internal degrees of freedom, such as isospin or colour, is given in chapter 6. We apply the above in a discussion of the classical scattering of particles off a 't Hooft-Polyakov monopole. In chapter 7 we elaborate on a Kaluza-Klein description of particles with internal degrees of freedom. The canonical formalism and the quantization of most of the preceeding systems are discussed in chapter 8. The dynamical systems given in chapters 3-7 are formulated on group manifolds. The procedure for obtaining the extension to super-group manifolds is briefly discussed in chapter 9. In chapter 10, we show that if a system admits only local Lagrangians for a configuration space Q, then under certain conditions, it admits a global Lagrangian when Q is enlarged to a suitable U(1) bundle over Q. Conditions under which a symplectic form is derivable from a Lagrangian are also found. (orig./HSI)

  18. The LP-FP-2 severe fuel damage scenario and discussion of the relative influence of the transient and reflood phases in affecting the final condition of the bundle

    The purpose of this paper is to review the evidence from the OECD LP-FP-2 experiment that a high temperature excursion occurred within the center fuel module (CFM) during the reflood portion of the test, was caused by rapid metal-water reaction. It is shown that this reflood scenario explains many perplexing observations from the experiment, in particular, the small amount of fission products and hydrogen transported to the blowdown suppression tank (BST) as compared with the larger quantities trapped within the primary coolant system (PCS). The timing and destruction of the CFM upper tie plate, as well as the transport of fuel debris to the top of this plate, are also explained. In general, all measurements, observations, and analyses of the LP-FP-2 data indicate that most of the CFM damage occurred during a relatively short period of time coincident with the reflood portion of the experiment. 4 refs., 6 figs

  19. Preliminary report: NIF laser bundle review

    As requested in the guidance memo 1, this committe determined whether there are compelling reasons to recommend a change from the NIF CDR baseline laser. The baseline bundle design based on a tradeoff between cost and technical risk, which is replicated four times to create the required 192 beams. The baseline amplifier design uses bottom loading 1x4 slab and flashlamp cassettes for amplifier maintenance and large vacuum enclosures (2.5m high x 7m wide in cross-section for each of the two spatial filters in each of the four bundles. The laser beams are arranged in two laser bays configured in a u-shape around the target area. The entire bundle review effort was performed in a very short time (six weeks) and with limited resources (15 personnel part-time). This should be compared to the effort that produced the CDR design (12 months, 50 to 100 personnel). This committee considered three alternate bundle configurations (2x2, 4x2, and 4x4 bundles), and evaluated each bundle against the baseline design using the seven requested issues in the guidance memo: Cost; schedule; performance risk; maintainability/operability; hardware failure cost exposure; activation; and design flexibility. The issues were reviewed to identify differences between each alternate bundle configuration and the baseline

  20. Prioritary omalous bundles on Hirzebruch surfaces

    Aprodu, Marian; Marchitan, Marius

    2016-01-01

    An irreducible algebraic stack is called unirational if there exists a surjective morphism, representable by algebraic spaces, from a rational variety to an open substack. We prove unirationality of the stack of prioritary omalous bundles on Hirzebruch surfaces, which implies also the unirationality of the moduli space of omalous H-stable bundles for any ample line bundle H on a Hirzebruch surface (compare with Costa and Miro-Ŕoig, 2002). To this end, we find an explicit description of the duals of omalous rank-two bundles with a vanishing condition in terms of monads. Since these bundles are prioritary, we conclude that the stack of prioritary omalous bundles on a Hirzebruch surface different from P1 ×P1 is dominated by an irreducible section of a Segre variety, and this linear section is rational (Ionescu, 2015). In the case of the space quadric, the stack has been explicitly described by N. Buchdahl. As a main tool we use Buchdahl's Beilinson-type spectral sequence. Monad descriptions of omalous bundles on hypersurfaces in P4, Calabi-Yau complete intersection, blowups of the projective plane and Segre varieties have been recently obtained by A.A. Henni and M. Jardim (Henni and Jardim, 2013), and monads on Hirzebruch surfaces have been applied in a different context in Bartocci et al. (2015).

  1. Singular hermitian metrics on vector bundles

    De Cataldo, M A A

    1997-01-01

    We introduce a notion of singular hermitian metrics (s.h.m.) for holomorphic vector bundles and define positivity in view of $L^2$-estimates. Associated with a suitably positive s.h.m. there is a (coherent) sheaf 0-th kernel of a certain $d''$-complex. We prove a vanishing theorem for the cohomology of this sheaf. All this generalizes to the case of higher rank known results of Nadel for the case of line bundles. We introduce a new semi-positivity notion, $t$-nefness, for vector bundles, establish some of its basic properties and prove that on curves it coincides with ordinary nefness. We particularize the results on s.h.m. to the case of vector bundles of the form $E=F \\otimes L$, where $F$ is a $t$-nef vector bundle and $L$ is a positive (in the sense of currents) line bundle. As applications we generalize to the higher rank case 1) Kawamata-Viehweg Vanishing Theorem, 2) the effective results concerning the global generation of jets for the adjoint to powers of ample line bundles, and 3) Matsusaka Big Theor...

  2. Deformations of the generalised Picard bundle

    Let X be a nonsingular algebraic curve of genus g ≥ 3, and let Mξ denote the moduli space of stable vector bundles of rank n ≥ 2 and degree d with fixed determinant ξ over X such that n and d are coprime. We assume that if g = 3 then n ≥ 4 and if g = 4 then n ≥ 3, and suppose further that n0, d0 are integers such that n0 ≥ 1 and nd0 + n0d > nn0(2g - 2). Let E be a semistable vector bundle over X of rank n0 and degree d0. The generalised Picard bundle Wξ(E) is by definition the vector bundle over Mξ defined by the direct image pMξ *(Uξ x pX*E) where Uξ is a universal vector bundle over X x Mξ. We obtain an inversion formula allowing us to recover E from Wξ(E) and show that the space of infinitesimal deformations of Wξ(E) is isomorphic to H1(X, End(E)). This construction gives a locally complete family of vector bundles over Mξ parametrised by the moduli space M(n0,d0) of stable bundles of rank n0 and degree d0 over X. If (n0,d0) = 1 and Wξ(E) is stable for all E is an element of M(n0,d0), the construction determines an isomorphism from M(n0,d0) to a connected component M0 of a moduli space of stable sheaves over Mξ. This applies in particular when n0 = 1, in which case M0 is isomorphic to the Jacobian J of X as a polarised variety. The paper as a whole is a generalisation of results of Kempf and Mukai on Picard bundles over J, and is also related to a paper of Tyurin on the geometry of moduli of vector bundles. (author)

  3. Predictions of Critical Heat Flux Using the ASSERT-PV Subchannel Code for a CANFLEX Variant Bundle

    The ASSERT-PV subchannel code developed by AECL has been applied as a design-assist tool to the advanced CANDU1 reactor fuel bundle. Based primarily on the CANFLEX2 fuel bundle, several geometry changes (such as element sizes and pitchcircle diameters of various element rings) were examined to optimize the dryout power and pressure-drop performances of the new fuel bundle. An experiment was performed to obtain dryout power measurements for verification of the ASSERT-PV code predictions. It was carried out using an electrically heated, Refrigerant-134a cooled, fuel bundle string simulator. The axial power profile of the simulator was uniform, while the radial power profile of the element rings was varied simulating profiles in bundles with various fuel compositions and burn-ups. Dryout power measurements are predicted closely using the ASSERT-PV code, particularly at low flows and low pressures, but are overpredicted at high flows and high pressures. The majority of data shows that dryout powers are underpredicted at low inlet-fluid temperatures but overpredicted at high inlet-fluid temperatures

  4. Geometry of quantum principal bundles, 1

    Durdevic, M

    1995-01-01

    A theory of principal bundles possessing quantum structure groups and classical base manifolds is presented. Structural analysis of such quantum principal bundles is performed. A differential calculus is constructed, combining differential forms on the base manifold with an appropriate differential calculus on the structure quantum group. Relations between the calculus on the group and the calculus on the bundle are investigated. A concept of (pseudo)tensoriality is formulated. The formalism of connections is developed. In particular, operators of horizontal projection, covariant derivative and curvature are constructed and analyzed. Generalizations of the first structure equation and of the Bianchi identity are found. Illustrative examples are presented.

  5. Weak equivalence classes of complex vector bundles

    Hông-Vân Lê

    2006-01-01

    For any complex vector bundle Ek of rank k over a manifold Mm with Chern classes ci Î H2i(Mm, Z) and any non-negative integers l1, . . ., lk we show the existence of a positive number p(m, k) and the existence of a complex vector bundle Êk over Mm whose Chern classes are p(m, k) × li × ci Î H2i(Mm, Z). We also discuss a version of this statement for holomorphic vector bundles over projective algebraic manifolds.

  6. Vector supersymmetry in the universal bundle

    We present a vector supersymmetry for Witten-type topological gauge theories, and examine its algebra in terms of a superconnection formalism. When covariant constraints on the supercurvature are chosen, a correspondence is established with the universal bundle construction of Atiyah and Singer. The vector supersymmetry represents a certain shift operator in the curvature of the universal bundle, and can be used to generate the hierarchy of observables in these theories. This formalism should lead to the construction of vector supergravity theories, and perhaps to the gravitational analogue of the universal bundle. (orig.)

  7. Safety analysis report of the irradiation test of Type-B bundle

    Lee, Choong Sung; Lim, I. C.; Lee, B. C.; Ryu, J. S.; Kim, H. R

    2000-06-01

    The HANARO fuel, U{sub 3}Si-A1, has been developed by AECL and tested in NRU reactor. In the course of the fuel qualification tests, only one case was performed under the higher power condition than maximum linear power which was expected in the design stage. The Korea regulatory body, KINS imposed that HANARO shall be operated at the power level less than 24MW which is 80% of the design full power until HANARO shows the repetitive performance of the fuel at the power condition abov e 112.8KW/m. To resolve this imposition, KAERI designed two types of special test bundles: two non-instrumented(Type-A) and one instrumented(Type-B) test bundles. Two Type-A bundles were irradiated in HANARO: one of them has finished PIE and the other is under PIE. Type-B bundle was loaded in the core during 1.32 day at 1996, but outstanding FIV(flow induced vibration) was observed at the pool top because of long guide tube attached to the top of the bundle. The successful installation of the chimney fastener to fix the guide tube resulted in conducting the irradiation test of Type-B bundle again. The test will start at mid- July, 2000. In order to safely do the Type-B irradiation test, the safety analysis for the nuclear, mechanical and thermal-hydraulic aspects was performed. The reactivity worth and the maximum 1 near power predicted by VENTURE are 6.3mk/k and 121.6kW/m, respectively. Thermal margins for normal and transient conditions using MATRA-h, are assessed to satisfy the safety criteria.

  8. Calculation of power distributions for experimental bundles in the NRU loops

    The NRU reactor is a large D2O tank filled with many different kinds of experimental and isotope-producing fuel and absorber assemblies in a vertical orientation. Some of the most important facilities are closed H2O loops, in which advanced CANDU fuel concepts can be irradiated in CANDU bundle formats. Six bundles are arranged linearly in a fuel string assembly, which can be installed in a loop test section. Each bundle can be different, and can contain different types of fuel in each ring of elements, so that different fuel concepts and power histories can be run in the same irradiation. The most important piece of information needed to design these irradiations and interpret the PIE results is the power history that each fuel segment experiences during irradiation. The initial power distribution under nominal full reactor power conditions must be calculated to determine if experimental requirements and thermalhydraulic limits can be met. The expected power variation during irradiation must also be known. Following irradiation, the actual power history experienced by each fuel segment must be calculated, to help assess fuel performance. Since only the total loop power-to-coolant is measured directly, it must be broken down into the fission powers of all the different fuel types, elements and segments that were irradiated together. This paper presents a review of the method currently used to achieve this devolution. It involves the use of the 2-D neutron transport code WIMS-AECL to determine the fission power distribution through each type of fuel bundle in the loop string as a function of burnup, and an estimate of the axial thermal flux shape in the moderator outside of the loop. These are input to an accounting code, BURFEL, which uses them to calculate the initial relative powers in all fuel segments in the loop, and sums them to the total loop power. In predictive mode, this is renormalized to the expected loop power at full reactor power, calculated

  9. Effect of Testing Conditions on Fibre-Bundle Tensile Properties Part Ⅰ: Sample Preparation, Bundle Mass and Fibre Alignment of Wool Bundles

    YU Wei-dong; YAN Hao-jing; Ron Postle; Yang Shouren

    2002-01-01

    Due to the effects of samples and testing conditions on fibre-bundle tensile behaviour, it is necessary to investigate the relationships between experimental factors and tensile properties for the fibre-bumdle tensile tester (TENSOR). The effects of bundle sample preparation, fibre bundle mass and fibre alignment have been tested. The experimental results indicated that (1) the low damage in combing and no free-end fibres in the cut bundle are most important for the sample preparation; (2) the reasonable bundle mass is 400- 700tex, but the tensile properties measured should bemodified with the bundle mass because a small amount of bundle mass causes the scatter results, while the larger is the bundle mass, the more difficult to comb fibres parallel and to clamp fibre evenly; and (3) the fibre irregular arrangement forms a slack bundle resulting in interaction between fibres, which will affect the reproducibility and accuracy of the tensile testing.

  10. Self-mapping degrees of torus bundles and torus semi-bundles

    Sun, Hongbin; Wang, Shicheng; Wu, Jianchun

    2010-01-01

    Each closed oriented 3-manifold $M$ is naturally associated with a set of integers $D(M)$, the degrees of all self-maps on $M$. $D(M)$ is determined for each torus bundle and torus semi-bundle $M$. The structure of torus semi-bundle is studied in detail. The paper is a part of a project to determine $D(M)$ for all 3-manifolds in Thurston's picture.

  11. Quantum Bundle Description of Quantum Projective Spaces

    Ó Buachalla, Réamonn

    2012-12-01

    We realise Heckenberger and Kolb's canonical calculus on quantum projective ( N - 1)-space C q [ C p N-1] as the restriction of a distinguished quotient of the standard bicovariant calculus for the quantum special unitary group C q [ SU N ]. We introduce a calculus on the quantum sphere C q [ S 2 N-1] in the same way. With respect to these choices of calculi, we present C q [ C p N-1] as the base space of two different quantum principal bundles, one with total space C q [ SU N ], and the other with total space C q [ S 2 N-1]. We go on to give C q [ C p N-1] the structure of a quantum framed manifold. More specifically, we describe the module of one-forms of Heckenberger and Kolb's calculus as an associated vector bundle to the principal bundle with total space C q [ SU N ]. Finally, we construct strong connections for both bundles.

  12. Twin tori for a new bundle divertor

    A new bundle divertor system using the straight stagnation axis in toroidal field together with the uniform field along the axis is discussed in detail. We call this type of divertor as the ''muffler divertor'' because of its shape. (author)

  13. Noncommutative principal bundles through twist deformation

    Aschieri, Paolo; Pagani, Chiara; Schenkel, Alexander

    2016-01-01

    We construct noncommutative principal bundles deforming principal bundles with a Drinfeld twist (2-cocycle). If the twist is associated with the structure group then we have a deformation of the fibers. If the twist is associated with the automorphism group of the principal bundle, then we obtain noncommutative deformations of the base space as well. Combining the two twist deformations we obtain noncommutative principal bundles with both noncommutative fibers and base space. More in general, the natural isomorphisms proving the equivalence of a closed monoidal category of modules and its twist related one are used to obtain new Hopf-Galois extensions as twists of Hopf-Galois extensions. A sheaf approach is also considered, and examples presented.

  14. High-resolution flow structure measurements in a rod bundle

    Flow behaviour inside a rod bundle has been an active research topic since the early days of the nuclear power industry. Of particular interest in previous studies have been topics such as flow mixing, two-phase flow structure and mapping of two-phase flow transitions. The optimisation of fuel element design can only be achieved by truly understanding the nature of flow. The ultimate goal in this research is to enhance the heat transfer and increase the critical heat flux, which would improve the fuel economy. A better understanding of the flow would also improve nuclear safety as departure from nucleate boiling (DNB) can be predicted more accurately. The motivation for the current project (SUBFLOW) was to increase knowledge of the complex flow phenomena inside a rod bundle. A dedicated sub-channel flow test facility was designed and constructed at the Paul Scherrer Institut (PSI), Villigen, Switzerland. An adiabatic test loop has an up-scaled (1:2.6) vertical fuel rod bundle model with a 4 × 4 geometry. For the very first time, the wire-mesh sensor measurement technique was implemented in a rod bundle as two 64×64 conductivity wire-mesh sensors were installed in the upper part of the test section. The measurement technique enables one to study single- and two-phase flow behaviour with high spatial and temporal resolution. The research topics addressed in this thesis cover a wide range of flow conditions with and without a spacer grid in a rod bundle. The experimental campaign was started by studying natural mixing of a passive scalar to characterise the development of turbulent diffusion in an injection sub-channel and, later on, cross-mixing between adjacent sub-channels. The results were also used in comparison with the in-house CFD code PSI-Boil that is being developed at PSI. The code could estimate the mixing inside the sub-channel and the transition to cross-mixing with a good accuracy. As a natural transition, the SUBFLOW experiments were continued by

  15. High-resolution flow structure measurements in a rod bundle

    Ylönen, A. T.

    2013-07-01

    Flow behaviour inside a rod bundle has been an active research topic since the early days of the nuclear power industry. Of particular interest in previous studies have been topics such as flow mixing, two-phase flow structure and mapping of two-phase flow transitions. The optimisation of fuel element design can only be achieved by truly understanding the nature of flow. The ultimate goal in this research is to enhance the heat transfer and increase the critical heat flux, which would improve the fuel economy. A better understanding of the flow would also improve nuclear safety as departure from nucleate boiling (DNB) can be predicted more accurately. The motivation for the current project (SUBFLOW) was to increase knowledge of the complex flow phenomena inside a rod bundle. A dedicated sub-channel flow test facility was designed and constructed at the Paul Scherrer Institut (PSI), Villigen, Switzerland. An adiabatic test loop has an up-scaled (1:2.6) vertical fuel rod bundle model with a 4 × 4 geometry. For the very first time, the wire-mesh sensor measurement technique was implemented in a rod bundle as two 64×64 conductivity wire-mesh sensors were installed in the upper part of the test section. The measurement technique enables one to study single- and two-phase flow behaviour with high spatial and temporal resolution. The research topics addressed in this thesis cover a wide range of flow conditions with and without a spacer grid in a rod bundle. The experimental campaign was started by studying natural mixing of a passive scalar to characterise the development of turbulent diffusion in an injection sub-channel and, later on, cross-mixing between adjacent sub-channels. The results were also used in comparison with the in-house CFD code PSI-Boil that is being developed at PSI. The code could estimate the mixing inside the sub-channel and the transition to cross-mixing with a good accuracy. As a natural transition, the SUBFLOW experiments were continued by

  16. Crosstalk analysis of carbon nanotube bundle interconnects

    Zhang, Kailiang; Tian, Bo; Zhu, Xiaosong; WANG, FANG; Wei, Jun

    2012-01-01

    Carbon nanotube (CNT) has been considered as an ideal interconnect material for replacing copper for future nanoscale IC technology due to its outstanding current carrying capability, thermal conductivity, and mechanical robustness. In this paper, crosstalk problems for single-walled carbon nanotube (SWCNT) bundle interconnects are investigated; the interconnect parameters for SWCNT bundle are calculated first, and then the equivalent circuit has been developed to perform the crosstalk analys...

  17. A Geometric Approach to Noncommutative Principal Bundles

    Wagner, Stefan

    2011-01-01

    From a geometrical point of view it is, so far, not sufficiently well understood what should be a "noncommutative principal bundle". Still, there is a well-developed abstract algebraic approach using the theory of Hopf algebras. An important handicap of this approach is the ignorance of topological and geometrical aspects. The aim of this thesis is to develop a geometrically oriented approach to the noncommutative geometry of principal bundles based on dynamical systems and the representation theory of the corresponding transformation group.

  18. Parahoric bundles on a compact Riemann surface

    Balaji, V

    2010-01-01

    Let $X$ be a compact Riemann surface of genus $g \\geq 2$. The aim of this paper is to study homomorphisms of certain discrete subgroups of $PSL(2, {\\mathbb R})$ into maximal compact subgroups of semisimple simply connected algebraic groups and relate them to torsors under a Bruhat-Tits group scheme. We also construct the moduli spaces of semistable parahoric bundles. These results generalize the theorem of Mehta and Seshadri on parabolic vector bundles.

  19. Hydrodynamic Experiments for a Flow Distribution of a 61-pin Wire-wrapped Rod Bundle

    Fuel assembly of the SFR (Sodium-cooled Fast breeder Reactor) type reactor generally has wire spacers which are wrapped around each fuel pin helically in axial direction. The configuration of a helical wire spacer guarantees the fuel rods integrity by providing the bundle rigidity, proper spacing between rods and promoting coolant mixing between subchannels. It is important to understand the flow characteristics in such a triangular array wire wrapped rod bundle in a hexagonal duct. The experimental work has been undertaken to quantify the friction and mixing parameters which characterize the flow distribution in subchannels for the KAERI's own bundle geometric configuration. This work presents the hydrodynamic experimental results for the flow distribution and the pressure drop in subchannels of a 61-pin wire wrapped rod bundle which has been fabricated considering the hydraulic similarity of the reference reactor. Hydrodynamic experiments for a 61-pin wire wrapped test assembly has been performed to provide the data of a flow distribution and pressure losses in subchannels for verifying the analysis capability of subchannel analysis codes for a KAERI's own prototype SFR reactor. Three type of sampling probes have been specially designed to conserve the shape of the flow area for each type of subchannels. All 126 subchannels have been measured to identify the characteristics of the flow distribution in a 37-pin rod assembly. Pressure drops at the interior and the edge subchannels have been also measured to recognize the friction losses of each type of subchannels

  20. Full conversion of materials and nuclear fuel research and test - TRIGA SSR 14 MW

    This article presents the HEU (high enrichment uranium) to LEU (low enrichment uranium) conversion of the TRIGA reactor at the Institute for Nuclear Research (Pitesti, Romania). This process began in 1992 when the first 4 LEU (23% U235 enrichment) fuel bundles integrated the reactor core in replacement of 4 HEU fuel bundles. By March 2004, the mixed reactor core had 18 LEU and 17 HEU fuel bundles by HEU-LEU replacement through successive steps of fueling. In 2006 the conversion process was completed and now we have a standard reactor core of 29 LEU fuel bundles

  1. Full conversion of materials and nuclear fuel research and test - TRIGA SSR 14 MW

    Ciocanescu, M.; Preda, M.; Iorgulis, C. [Triga Reactor, Instittute For Nuclear Research Pitesti (Romania)

    2007-07-01

    This article presents the HEU (high enrichment uranium) to LEU (low enrichment uranium) conversion of the TRIGA reactor at the Institute for Nuclear Research (Pitesti, Romania). This process began in 1992 when the first 4 LEU (23% U235 enrichment) fuel bundles integrated the reactor core in replacement of 4 HEU fuel bundles. By March 2004, the mixed reactor core had 18 LEU and 17 HEU fuel bundles by HEU-LEU replacement through successive steps of fueling. In 2006 the conversion process was completed and now we have a standard reactor core of 29 LEU fuel bundles.

  2. Luncheon address: Early days of CANDU fuel

    I will briefly describe how the original dimensions of the fuel bundle were defined and how that early designs of fuel evolved. I will also touch on some of the historical events of the materials and experiments which effected the fuel programme. Also how I became with Canada's Nuclear Fuel programme. (author)

  3. K-Theories for Certain Infinite Rank Bundles

    Larrain-Hubach, Andres

    2011-01-01

    Several authors have recently constructed characteristic classes for classes of infinite rank vector bundles appearing in topology and physics. These include the tangent bundle to the space of maps between closed manifolds, the infinite rank bundles in the families index theorem, and bundles with pseudodifferential operators as structure group. In this paper, we construct the corresponding K-theories for these types of bundles. We develop the formalism of these theories and use their Chern ch...

  4. Effect of left bundle branch block on TIMI frame count

    Hatice Tolunay; Ahmet Kasapkara; İsa Öner Yüksel; Nurcan Başar; Ayşe Saatcı Yaşar; Mehmet Bilge

    2010-01-01

    Aim: Left bundle branch block is an independent risk factorfor cardiac mortality. In this study we aimed to evaluatecoronary blood flow with TIMI frame count in patients with left bundle branch block and angiographically proven normal coronary arteries.Materials and methods: We retrospectively studied 17 patients with left bundle branch block and as a control group 16 patients without left bundle branch block. All patientshad angiographically proven normal coronary arteries.Left bundle branch...

  5. Product-bundling and Incentives for Merger and Strategic Alliance

    Sue Mialon

    2009-01-01

    This paper analyzes firms' choice between a merger and a strategic alliance in bundling their product with other complementary products. We consider a framework in which firms can improve profits only from product-bundling. While mixed bundling is not profitable, pure bundling is because pure bundling reduces consumers' choices, and thus softens competition among firms. Firms benefit the most from this reduced competition if they form an alliance. Firms do not gain as much from a merger becau...

  6. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

    This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized. Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified

  7. Analysis of steady state combined forced and free convection data in rod bundles

    Fuel and blanket assemblies in an LMFBR are subjected to a wide range of power and power gradients during their life in the reactor. To accommodate these changes the assemblies operate in a wide range of flow regimes extending from forced convection, turbulent flow, to mixed convection, laminar flow. At low flow conditions the transverse temperature gradient in an assembly is considerably flattened because of energy redistribution by not only wire-wrap mixing and thermal conduction but also by flow redistribution because of buoyancy-induced crossflow. This has significance in LMFBR design. For the mixed convection regime of bundle operation, the transverse velocity profiles within a bundle change axially because of buoyancy-induced crossflow. It was therefore decided to use the ENERGY II and ENERGY III computer programs for the analysis of the rod bundle mixed convection data

  8. Mechanism of Actin Filament Bundling by Fascin

    Jansen, Silvia; Collins, Agnieszka; Yang, Changsong; Rebowski, Grzegorz; Svitkina, Tatyana; Dominguez, Roberto (UPENN); (UPENN-MED)

    2013-03-07

    Fascin is the main actin filament bundling protein in filopodia. Because of the important role filopodia play in cell migration, fascin is emerging as a major target for cancer drug discovery. However, an understanding of the mechanism of bundle formation by fascin is critically lacking. Fascin consists of four {beta}-trefoil domains. Here, we show that fascin contains two major actin-binding sites, coinciding with regions of high sequence conservation in {beta}-trefoil domains 1 and 3. The site in {beta}-trefoil-1 is located near the binding site of the fascin inhibitor macroketone and comprises residue Ser-39, whose phosphorylation by protein kinase C down-regulates actin bundling and formation of filopodia. The site in {beta}-trefoil-3 is related by pseudo-2-fold symmetry to that in {beta}-trefoil-1. The two sites are {approx}5 nm apart, resulting in a distance between actin filaments in the bundle of {approx}8.1 nm. Residue mutations in both sites disrupt bundle formation in vitro as assessed by co-sedimentation with actin and electron microscopy and severely impair formation of filopodia in cells as determined by rescue experiments in fascin-depleted cells. Mutations of other areas of the fascin surface also affect actin bundling and formation of filopodia albeit to a lesser extent, suggesting that, in addition to the two major actin-binding sites, fascin makes secondary contacts with other filaments in the bundle. In a high resolution crystal structure of fascin, molecules of glycerol and polyethylene glycol are bound in pockets located within the two major actin-binding sites. These molecules could guide the rational design of new anticancer fascin inhibitors.

  9. An experimental study of water distribution from a jet to a single pin and pin bundles

    A knowledge of the water distribution in spray cooling of overheated water reactor fuel bundles is necessary for the proper analysis of heat removal under accident conditions. Results are presented of experiments on the distribution of water from a single horizontal jet to a cold vertical pin and both heated and unheated pin bundles. The flow running down a single cold pin has been determined for a range of jet impact angles, jet flow rates, and jet diameters. This is a critical flowrate above which water detaches from the pin at normal impact and it has been shown that at this critical flowrate the fraction of water flowing down the pin is proportional to the cosine of the impact angle. Tests with a cold six-pin sector of a three-ringed 36 pin bundle showed that distribution of water to the pins was non-uniform and sensitive to both jet orientation and velocity, as may be expected from the single pin results. Experiments on a hot six-pin showed that individual pins quenched at a rate predictable from cold six-pin flow distribution tests and a hot single pin 'calibration' test, once wetting was fully established at the water injection level. Observations made in a 36 pin heated bundle with six water jets confirmed the findings on the smaller bundles. It has also been shown that during delays in the establishment of wetting, the water distribution will be markedly different to that once wetting is established. (author)

  10. Barrier fuel demonstrated at Quad Cities 2

    The article describes the nuclear design of the 144 zirconium barrier bundles inserted in the Quad Cities 2 core in December 1981, their locations and power history during their first cycle of irradiation, and the resulting linear heat generation rates (LHGRs) and fuel rod burnups when 16 of these zirconium barrier bundles were power ramped in March 1983. (author)

  11. NIF laser bundle review. Final report

    We performed additional bundle review effort subsequent to the completion of the preliminary report and are revising our original recommendations. We now recommend that the NIF baseline laser bundle size be changed to the 4x2 bundle configuration. There are several 4x2 bundle configurations that could be constructed at a cost similar to that of the baseline 4x12 (from $11M more to about $11M less than the baseline; unescalated, no contingency) and provide significant system improvements. We recommend that the building cost estimates (particularly for the in-line building options) be verified by an architect/engineer (A/E) firm knowledgeable about building design. If our cost estimates of the in-line building are accurate and therefore result in a change from the baseline U-shaped building layout, the acceptability of the in-line configuration must be reviewed from an operations viewpoint. We recommend that installation, operation, and maintenance of all laser components be reviewed to better determine the necessity of aisles, which add to the building cost significantly. The need for beam expansion must also be determined since it affects the type of bundle packing that can be used and increases the minimum laser bay width. The U-turn laser architecture (if proven viable) offers a reduction in building costs since this laser design is shorter than the baseline switched design and requires a shorter laser bay

  12. Nuclear fuel assembly

    A nuclear fuel assembly includes and upper yoke, a base, an elongated, outer flow channel disposed substantially along the entire length of the fuel assembly and an elongated, internal, central water cross, formed by four, elongated metal angles, that divides the nuclear fuel assembly into four, separate, elongated fuel sections and that provides a centrally disposed path for the flow of subcooled neutron moderator along the length of the fuel assembly. A separate fuel bundle is located in each of the four fuel sections and includes an upper tie plate, a lower tie plate and a plurality of elongated fuel rods disposed therebetween. Preferably, each upper tie plate is formed from a plurality of interconnected thin metal bars and includes an elongated, axially extending pin that is received by the upper yoke of the fuel assembly for restraining lateral motion of the fuel bundle while permitting axial movement of the fuel bundle with respect to the outer flow channel. The outer flow channel is fixedly secured at its opposite longitudinal ends to the upper yoke and to the base to permit the fuel assembly to be lifted and handled in a vertical position without placing lifting loads or stresses on the fuel rods. The yoke, removably attached at the upper end of the fuel assembly to four structural ribs secured to the inner walls of the outer flow channel, includes, as integrally formed components, a lifting bail or handle, laterally extending bumpers, a mounting post for a spring assembly, four elongated apertures for receiving with a slip fit the axially extending pins mounted on the upper tie plates and slots for receiving the structural ribs secured to the outer flow channel. Locking pins securely attach the yoke to the structural ribs enabling the fuel assembly to be lifted as an entity

  13. Severe Fuel-Damage Scoping Test post-irradiation examination results

    The fuel bundle from the Severe Fuel Damage Scoping Test, conducted in the Power Burst Facility as part of the international Severe Fuel Damage Research Program, was examined posttest. This paper presents the results of the nondestructive portion of the examination, including gross gamma scanning, neutron radiography, and tomographic reconstruction of cross sections through the bundle using the neutron radiographs

  14. Sertification of fuel cladding and grids materials in out of pile conditions

    The basic standard specifications for fuel rod cladding and bundle materials, are selected. In this paper the standard specifications of material for Zircaloy and plugs and stainless steel springs of fuel rod cladding are presented. The material specification for a Zircaloy fuel bundle assembly Cgrids) is also given. (author)

  15. Fuel management simulations for 0.9% SEU in CANDU 6 reactors

    Slightly Enriched Uranium (SEU) of 0.9 weight % 235U enrichment is a promising fuel cycle option for CANDU reactors. An important component of the investigation of this option is the demonstration of the feasibility of on-line refuelling with this fuel type in reactor physics fuel-management simulations. Two fuel-management schemes have been investigated in detail during 500-day core-follow simulations, these were a 2-bundle-shift and a 4-bundle-shift axial refuelling scheme. The 43-element CANFLEX fuel design has been used in these studies because of its improved fuel performance characteristics in this application. The results of the studies are discussed in detail in this paper. The most significant conclusion of this study was that both 2- and 4-bundle-shift refuelling schemes with CANFLEX fuel result in bundle power and bundle power boost envelopes that meet current fuel-performance requirements. (author)

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

    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

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

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

    2014-10-15

    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.

  18. Turbulent flow through two asymmetric rod bundles

    Measurements of the mean velocity, of the wall shear stresses, and of the turbulence have been performed in four wall subchannels of rod bundles of four parallel rods enclosed in a rectangular channel. The pitch-to-diameter ratio was P/D=1.148 and the wall-to-diameter ratios ranged from 1.045 to 1.252. The full Reynolds stress tensor has been determined by hot-wire technique. The results of the turbulences intensities show that the flow through rod bundles differs widely from flow through circular tubes. More sophisticated analytical tools than presently available are required to predict turbulent flow through rod bundles with sufficient accuracy

  19. Explicit core-follow simulations for a CANDU 6 reactor fuelled with recovered-uranium CANFLEX bundles

    Recovered uranium (RU) is a by-product of many light-water reactor (LWR) fuel recycling programs. After fission products and plutonium (Pu) have been removed from spent LWR fuel, RU is left. A fissile content in the RU of 0.9 to 1.0% makes it impossible for reuse in an LWR without re-enrichment, but CANDU reactors have a sufficiently high neutron economy to use RU as fuel. Explicit core-follow simulations were run to analyse the viability of RU as a fuel for existing CANDU 6 cores. The core follow was performed with RFSP, using WIMS-AECL lattice properties. During the core follow, channel powers and bundle powers were tracked to determine the operating envelope for RU in a CANFLEX bundle. The results show that RU fits the operating criteria of a generic CANDU 6 core and is a viable fuel option in CANDU reactors. (author)

  20. Fuel assembly

    Fuel rods are arranged in a lattice-like structure by way of a plurality of spacers and the lower ends thereof are fixed to a lower tie plate for assembling a fuel rod bundle. The outer circumference is surrounded by a basket having a plurality of openings and the basket is surrounded by a channel box. The basket is connected to a handle at the upper end and to a lower tie plate at the lower end and, further, defined with a scraper at each of openings. Coolants flown from the lower tie plate to the channel box flow the channels between the channel box and the basket and a fuel rod bundle, uprise while forming a two-phase flow and flow out from the upper end of the channel box. Since no upper tie plate is present, pressure loss of coolants flow is reduced, and liquid membranes of coolants are peeled off by the scraper disposed at the opening of the basket, which contributes to the improvement of the limit power. In addition, fuel rods are inspected and cleaned easily. (N.H.)

  1. Fuel Temperature Characteristics for Fuel Channels using Burnable Poison in the CANDU reactor

    Although the CANFLEX RU fuel bundle loaded 11.0 wt% Er2O3 are originally designed focused on the safety characteristics, the fuel temperature characteristics is revealed to be not deteriorated but rather is slightly enhanced by the decreased fuel temperature in the outer ring compared with that of standard 37 fuel bundle. Recently, for an equilibrium CANDU core, the power coefficient was reported to be slightly positive when newly developed Industry Standard Tool set reactor physics codes were used. Therefore, it is required to find a new way to effectively decrease the positive power coefficient of CANDU reactor without seriously compromising the economy. In order to make the power coefficient of the CANDU reactor negative at the operating power, Roh et al. have evaluated the various burnable poison (BP) materials and its loading scheme in terms of the fuel performance and reactor safety characteristics. It was shown that reactor safety characteristics can be greatly improved by the use of the BP in the CANDU reactor. In a view of safety, the fuel temperature coefficient (FTC) is an important safety parameter and it is dependent on the fuel temperature. For an accurate evaluation of the safety-related physics parameters including FTC, the fuel temperature distribution and its correlation with the coolant temperature should be accurately identified. Therefore, we have evaluated the fuel temperature distribution of a CANFLEX fuel bundle loaded with a burnable poison and compared the standard 37 element fuel bundle and CANFELX-NU fuel bundle

  2. Porous Silicon and Denim Fiber Bundle Characterization

    Deuro, Randi Ellen

    My thesis research aims to characterize and exploit materials in an efficient, rapid, non-destructive manner. Part I of this document summarizes my research on porous silicon (pSi) design, fabrication, and surface modification for use as a novel chemical sensor. The optimization of fabrication process parameters (etching time, etching solution, electrode shape, and the fixing process) on pSi photoluminescence (PL) is presented. I have also investigated the effects of analyte vapors (acetonitrile, toluene, methanol, acetone) on the pSi PL and surface chemistry using luminescence and Fourier-transform infrared (FT-IR) spectroscopy and microscopy methods. The mechanism and benefits of one method of pSi surface modification and protection (ultraviolet (UV) hydrosilylation) will also be presented. Finally, high thorough-put methods of pSi sensor production are described. In Part II of this document, I introduce a novel technique for analyzing and discriminating among denim fiber bundles. An investigation into the benefits of luminescence-based multispectral imaging (LMSI) for denim fiber bundle identification has been conducted. I explore the power of nitromethane (CH 3NO2) based quenching in fiber bundle classification and identify the quenching mechanism. The luminescence spectra (450 - 850 nm) and images from the denim fiber bundles were obtained while exciting at 325 nm or 405 nm. Here, LMSI data were recorded in < 10 s and subsequently assessed by principal component analysis (PCA) and rendered red, green, blue (RGB) component histograms. The results show that LMSI data can be used to rapidly and uniquely classify all the fiber bundle types studied in this research. These non-destructive techniques eliminate extensive sample preparation and allow for rapid multispectral image collection, analysis, and assessment. The quenching data also revealed that the dye molecules within the individual fiber bundles exhibited dramatically different accessibilities to CH 3NO2.

  3. Burn up Analysis for Fuel Assembly Unit i n a Pressurized Heavy Water CANDU Reactor

    MCNPX code has been used for modeling a nd simulation of an assembly of CANDU Fuel bundle . The assembly is composed of a heterogeneous lattice of 37-element natural Uranium fuel, heavy water moderator and coolant. The fuel bundle is burnt in normal operation conditions of CANDU reactors. The effective multiplication factor (Keff ) of the bundle is calculated as a function of fuel burnup. The flux and power distribution are determined. Comparing t he concentrations of both Uranium and Plutonium isotopes are analyzed in the bundle. The results of the present model with the results of a benchmark problem, a good agreement was found PWR

  4. Numerical simulation of flow-induced vibrations in tube bundles

    Full text of publication follows: In many industrial components mechanical structures like rod cluster control assembly, fuel assembly and heat exchanger tube bundles are submitted to complex flows causing possible vibrations and damage. Fluid forces are usually split into two parts: structure motion independent forces and fluid-elastic forces coupled with tube motion and responsible for possible dynamic instability development leading to possible short term failures through high amplitude vibrations. Most classical fluid force identification methods rely on structure response experimental measurements associated with convenient data processes. Owing to recent improvements in Computational Fluid Dynamics (C.F.D.), numerical fluid force identification is now practicable in the presence of industrial configurations. The present paper is devoted to numerical simulation of flow-induced vibrations of tube bundles submitted to single-phase cross flows by using C.F.D. codes. Direct Numerical Simulation (D.N.S.), Arbitrary Lagrange Euler formulation (A.L.E.) and code coupling process are involved to predict fluid forces responsible for tube bundle vibrations in the presence of fluid structure and fluid-elastic coupling effects. In the presence of strong multi-physics coupling, simulation of flow-induced vibrations requires a fluid structure code coupling process. The methodology consists in solving in the same time thermohydraulics and mechanics problems by using an A.L.E. formulation for the fluid computation. The purpose is to take into account coupling between flow and structure motions in order to be able to capture coupling effects. From a numerical point of view, there are three steps in the computation: the fluid problem is solved on the computational domain; fluid forces acting on the moving tube are estimated; finally they are introduced in the structure solver providing the tube displacement that is used to actualize the fluid computational domain. Specific

  5. Bundling in semiflexible polymers: A theoretical overview.

    Benetatos, Panayotis; Jho, YongSeok

    2016-06-01

    Supramolecular assemblies of polymers are key modules to sustain the structure of cells and their function. The main elements of these assemblies are charged semiflexible polymers (polyelectrolytes) generally interacting via a long(er)-range repulsion and a short(er)-range attraction. The most common supramolecular structure formed by these polymers is the bundle. In the present paper, we critically review some recent theoretical and computational advances on the problem of bundle formation, and point a few promising directions for future work. PMID:26813628

  6. A bundle of sticks in my garden

    Farran, Sue

    2012-01-01

    The English law of property is often described as a ‘bundle of sticks’ in which each ‘stick’ represents a particular right. Gardens challenge these rights and wreak havoc on the ‘bundle of sticks’. This paper looks at the twenty-first century manifestations of community engagement with ground and explores how ‘gardening’ is undermining concepts of ownership, possession and management of land and how the fence between what is private and what is public is being encroached and challenged by com...

  7. Characteristic classes of quantum principal bundles

    Durdevic, M

    1995-01-01

    A noncommutative-geometric generalization of classical Weil theory of characteristic classes is presented, in the conceptual framework of quantum principal bundles. A particular care is given to the case when the bundle does not admit regular connections. A cohomological description of the domain of the Weil homomorphism is given. Relations between universal characteristic classes for the regular and the general case are analyzed. In analogy with classical geometry, a natural spectral sequence is introduced and investigated. The appropriate counterpart of the Chern character is constructed, for structures admitting regular connections. Illustrative examples and constructions are presented.

  8. Scaling Shift in Multicracked Fiber Bundles

    Manca, Fabio; Giordano, Stefano; Palla, Pier Luca; Cleri, Fabrizio

    2014-12-01

    Bundles of fibers, wires, or filaments are ubiquitous structures in both natural and artificial materials. We investigate the bundle degradation induced by an external damaging action through a theoretical model describing an assembly of parallel fibers, progressively damaged by a random population of cracks. Fibers in our model interact by means of a lateral linear coupling, thus retaining structural integrity even after substantial damage. Monte Carlo simulations of the Young's modulus degradation for increasing crack density demonstrate a remarkable scaling shift between an exponential and a power-law regime. Analytical solutions of the model confirm this behavior, and provide a thorough understanding of the underlying physics.

  9. Safe Harbors for Quantity Discounts and Bundling

    Dennis W. Carlton; Michael Waldman

    2008-01-01

    The courts and analysts continue to struggle to articulate safe harbors for a wide variety of common business pricing practices in which either a single product is sold at a discount if purchased in bulk or in which multiple products are bundled together at prices different from the ones that would emerge if the products were purchased separately. The phenomenon of tying in which the sale of one product is conditioned on the purchase of another is closely related to bundling. Its analysis rel...

  10. Improved fuel rod support means

    A fuel bundle for a nuclear reactor having a plurality of fuel rods supported between spaced tie plates, wherein coolant flows through said tie plates and past said fuel rods, characterized by: an end plug disposed between an end of each fuel rod and the adjacent tie plate, and means defining a passage for the flow of coolant through the interface between said end plug and said tie plates to minimize crud buildup at said interface

  11. Bundle 13 position verification tool description and on-reactor use

    To address the Power Pulse problem, Bruce B uses Gap: a comprehensive monitoring program by the station to maintain the gap between the fuel string and the upstream shield plug. The gap must be maintained within a band. The gap must not be so large as to allow excessive reactivity increases or cause high impact forces during reverse flow events. It should also not be so small as to cause crushed fuel during rapid, differential reactor/fuel string cool downs. Rapid cool downs are infrequent. The Bundle 13 Position Verification Tool (BPV tool) role is to independently measure the position of the upstream bundle of the fuel string. The measurements are made on-reactor, on-power and will allow verification of the Gap Management system's calculated fuel string position. This paper reviews the reasons for developing the BPV tool. Design issues relevant to safe operation in the fuelling machine, fuel channel and fuel handling equipment are also reviewed. Tests ensuring no adverse effects on channel pressure losses are described and actual on-reactor, on-power results are discussed. (author). 4 figs

  12. Impact of bundle deformation on CHF: ASSERT-PV assessment of extended burnup Bruce B bundle G85159W

    This paper presents a subchannel thermalhydraulic analysis of the effect on critical heat flux (CHF) of bundle deformation such as element bow and diametral creep. The bundle geometry is based on the post-irradiation examination (PIE) data of a single bundle from the Bruce B Nuclear Generating Station, Bruce B bundle G85159W, which was irradiated for more than two years in the core during reactor commissioning. The subchannel code ASSERT-PV IST is used to assess changes in CHF and dryout power due to bundle deformation, compared to the reference, undeformed bundle. (author)

  13. Korea's CANDU fuel R and D program

    As the first R and D activity led to the nuclear fuel industrialization in Korea, KAERI had successfully developed the CANDU-6 fuel bundle in the period of 1981 to 1986 and has commercially produced more than 35,000 fuel bundles for the use in Wolsong Unit 1 since 1987. The commercial production of the CANDU-6 fuel in KAERI will be terminated on the end of 1997 and KNFC will take over the mission of CANDU-6 fuel production with a capacity of 400 tons of uranium per year form 1998. (author)

  14. The advanced carrier bundle - comprehensive irradiation of materials in CANDU power reactors

    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)

  15. Abelian conformal field theory and determinant bundles

    Andersen, Jørgen Ellegaard; Ueno, K.

    2007-01-01

    Following [10], we study a so-called bc-ghost system of zero conformal dimension from the viewpoint of [14, 16]. We show that the ghost vacua construction results in holomorphic line bundles with connections over holomorphic families of curves. We prove that the curvature of these connections are...

  16. Optimization of a bundle divertor for FED

    Optimal double-T bundle divertor configurations have been obtained for the Fusion Engineering Device (FED). On-axis ripple is minimized, while satisfying a series of engineering constraints. The ensuing non-linear optimization problem is solved via a sequence of quadratic programming subproblems, using the VMCON algorithm. The resulting divertor designs are substantially improved over previous configurations

  17. Capacity efficiency of recovery request bundling

    Ruepp, Sarah Renée; Dittmann, Lars; Berger, Michael Stübert; Stidsen, Thomas Riis; Lagakos, Stephen; Perlovsky, Leonid; Jha, Manoi; Covaci, Brindusa; Zaharim, Azarni; Mastorakis, Nikos

    2010-01-01

    This paper presents a comparison of recovery methods in terms of capacity efficiency. In particular, a method where recovery requests are bundled towards the destination (Shortcut Span Protection) is evaluated against traditional recovery methods. Our simulation results show that Shortcut Span Pr...... Protection uses more capacity than the unbundled related methods, but this is compensated by easier control and management of the recovery actions....

  18. Line bundles on moduli and related spaces

    Huebschmann, Johannes

    2009-01-01

    Let G be a Lie goup, let M and N be smooth connected G-manifolds, let f be a smooth G-map from M to N, and let P denote the fiber of f. Given a closed and equivariantly closed relative 2-form for f with integral periods, we construct the principal G-circle bundles with connection on P having the given relative 2-form as curvature. Given a compact Lie group K, a biinvariant Riemannian metric on K, and a closed Riemann surface S of genus s, when we apply the construction to the particular case where f is the familiar relator map from a product of 2s copies of K to K we obtain the principal K-circle bundles on the associated extended moduli spaces which, via reduction, then yield the corresponding line bundles on possibly twisted moduli spaces of representations of the fundamental group of S in K, in particular, on moduli spaces of semistable holomorphic vector bundles or, more precisely, on a smooth open stratum when the moduli space is not smooth. The construction also yields an alternative geometric object, d...

  19. Bundle Gerbes Applied to Quantum Field Theory

    Carey, A L; Murray, M; Carey, Alan; Mickelsson, Jouko; Murray, Michael

    2000-01-01

    This paper reviews recent work on a new geometric object called a bundle gerbe and discusses some new examples arising in quantum field theory. One application is to an Atiyah-Patodi-Singer index theory construction of the bundle of fermionic Fock spaces parametrized by vector potentials in odd space dimensions and a proof that this leads in a simple manner to the known Schwinger terms (Mickelsson-Faddeev cocycle) for the gauge group action. This gives an explicit computation of the Dixmier-Douady class of the associated bundle gerbe. The method works also in other cases of fermions in external fields (external gravitational field, for example) provided that the APS theorem can be applied; however, we have worked out the details only in the case of vector potentials. Another example, in which the bundle gerbe curvature plays a role, arises from the WZW model on Riemann surfaces. A further example is the `existence of string structures' question. We conclude by showing how global Hamiltonian anomalies fit with...

  20. Quantum field theories on Hilbert bundles

    We investigate whether it is possible to maintain the computational features of QED while avoiding some of its mathematical difficulties by formulating QFTs on Hilber bundles. This encounters two problems: 1) Haag's theorem persists, and 2) admissible fields do not generate motions on the base space. To do the latter, the coupling constant has to be a vector field upon the base space. (orig.)

  1. Capacity efficiency of recovery request bundling

    Ruepp, Sarah Renée; Dittmann, Lars; Berger, Michael Stübert; Stidsen, Thomas Riis; Lagakos, Stephen; Perlovsky, Leonid; Jha, Manoi; Covaci, Brindusa; Zaharim, Azarni; Mastorakis, Nikos

    2010-01-01

    This paper presents a comparison of recovery methods in terms of capacity efficiency. In particular, a method where recovery requests are bundled towards the destination (Shortcut Span Protection) is evaluated against traditional recovery methods. Our simulation results show that Shortcut Span...

  2. Riemann Surfaces: Vector Bundles, Physics, and Dynamics

    Sikander, Shehryar

    the monodromy with respect to the pulled back connection. The formula for the representation includes a series with coefficients as iterated integrals. This series is closely related to the cyclotomic version of the Drinfel'd associator. The geodesic flow in the unit the tangent bundle of this Teichmueller...

  3. Feasibility of Accident-Tolerant FCM Replacement Fuel for CANDUs

    For enhanced accident tolerance, an innovative fuel concept, the fully ceramic microencapsulated (FCM) fuel based on the particle fuel concept of a gas-cooled reactor, is proposed to replace the conventional UO2 fuel bundle of existing and advanced CANDU reactors. In this study, the feasibility of replacing conventional UO2 fuel bundle with the accident-tolerant FCM fuel bundle has been assessed in view of core neutronics compatibility, accident-tolerance, and fuel cycle management. From the study, it was demonstrated that the FCM replacement fuel can provide resolution to CANDU generic issues by ensuring not only enhanced accident tolerance, but also an improved fuel cycle management. The accident-tolerant FCM fuel concept is proposed for replacing the conventional UO2 fuel bundle in CANDUs. The FCM fuel is shown to be neutronically compatible with existing core and the core residence time can be increased by more than 100 days. Accident-tolerance is remarkably enhanced by key features of the FCM fuel: it is refractory, thermo-mechanically and chemically stable, and fission product retentive. Less fuel feed and discharge obtained with the FCM fuel provide large savings in the spent fuel management burden charge and reduces the burden to the spent fuel storage facility in the long run. The smaller amount of minor actinides in the discharge bundles, together with the fission product retention and corrosion resistant features of the FCM fuel, should facilitate the long-term dry disposals of the spent fuel. From this study, it has been demonstrated that the CANDU FCM fuel is a feasible and viable option for CANDU reactors. The technology readiness level of the FCM fuel design and manufacturing is close to a lead test bundle loading for near-term deployment

  4. Experimental study of new generation WWER-1000 fuel assemblies at JSC NCCP

    An experimental program for the study of fuel assembly thermomechanical stability has been established together with RF SSC IPPE and Russian Scientific Center Kurchatov Institute. Assembly fragments and small dummy models of fuel assembly skeletons and fuel rod bundles have been used for the tests. The test results are used for the design selection, verification of the design codes and substantiation of operating capacity of fuel assemblies with a rigid skeleton. The mechanical characteristics of units make it possible to perform fuel assembly strength and rigidity calculations, including the cases of abnormal operation. The mechanical characteristics of the skeleton and fuel rod bundle dummy models make it possible to check for the adequacy of the fuel assembly design model. The mechanical characteristics obtained during fuel rods bundle push through experiments make it possible to substantiate the fuel assembly serviceability under the conditions of fuel rods bundle and skeleton interaction

  5. Models and characteristics of interchannel exchange in pin bundles cooled by liquid metal

    Experimental results on convective and turbulence mass, momentum and energy exchange in pin bundle cooled by liquid metal obtained by electromagnetic and thermal track techniques are generalized. The basis for analytical models of convective, turbulence exchange by momentum and energy, as well as heat transport due to fuel pin heat conduction are presented. Correlations derived are analyzed in comparison with the other authors' data. An influence of interchannel exchange on coolant and pin temperature distributions is illustrated by some examples. (author)

  6. Holomorphic Vector Bundle on Hopf Manifolds with Abelian Fundamental Groups

    Xiang Yu ZHOU; Wei Ming LIU

    2004-01-01

    Let X be a Hopf manifolds with an Abelian fundamental group. E is a holomorphic vector bundle of rank r with trivial pull-back to W = Cn - {0}. We prove the existence of a non-vanishing section of L(×) E for some line bundle on X and study the vector bundles filtration structure of E. These generalize the results of D. Mall about structure theorem of such a vector bundle E.

  7. Anatomic Double-Bundle Posterior Cruciate Ligament Reconstruction

    Chahla, Jorge; Nitri, Marco; Civitarese, David; Dean, Chase S.; Moulton, Samuel G.; LaPrade, Robert F.

    2016-01-01

    The posterior cruciate ligament (PCL) is known to be the main posterior stabilizer of the knee. Anatomic single-bundle PCL reconstruction, focusing on reconstruction of the larger anterolateral bundle, is the most commonly performed procedure. Because of the residual posterior and rotational tibial instability after the single-bundle procedure and the inability to restore the normal knee kinematics, an anatomic double-bundle PCL reconstruction has been proposed in an effort to re-create the n...

  8. Existence of vector bundles and global resolutions for singular surfaces

    Vezzosi, G; S. SCHROER

    2002-01-01

    Abstract- We prove two results about vector bundles on singular algebraic surfaces. First, on proper surfaces there are vector bundles of rank two with arbitrarily large second Chern number and fixed determinant. Second, on separated normal surfaces any coherent sheaf is the quotient of a vector bundle. As a consequence, for such surfaces the Quillen K-theory of vector bundles coincides with the Waldhausen K-theory of perfect complexes. Examples show that, on non-separated schemes, usually...

  9. Interplanetary Overlay Network Bundle Protocol Implementation

    Burleigh, Scott C.

    2011-01-01

    The Interplanetary Overlay Network (ION) system's BP package, an implementation of the Delay-Tolerant Networking (DTN) Bundle Protocol (BP) and supporting services, has been specifically designed to be suitable for use on deep-space robotic vehicles. Although the ION BP implementation is unique in its use of zero-copy objects for high performance, and in its use of resource-sensitive rate control, it is fully interoperable with other implementations of the BP specification (Internet RFC 5050). The ION BP implementation is built using the same software infrastructure that underlies the implementation of the CCSDS (Consultative Committee for Space Data Systems) File Delivery Protocol (CFDP) built into the flight software of Deep Impact. It is designed to minimize resource consumption, while maximizing operational robustness. For example, no dynamic allocation of system memory is required. Like all the other ION packages, ION's BP implementation is designed to port readily between Linux and Solaris (for easy development and for ground system operations) and VxWorks (for flight systems operations). The exact same source code is exercised in both environments. Initially included in the ION BP implementations are the following: libraries of functions used in constructing bundle forwarders and convergence-layer (CL) input and output adapters; a simple prototype bundle forwarder and associated CL adapters designed to run over an IPbased local area network; administrative tools for managing a simple DTN infrastructure built from these components; a background daemon process that silently destroys bundles whose time-to-live intervals have expired; a library of functions exposed to applications, enabling them to issue and receive data encapsulated in DTN bundles; and some simple applications that can be used for system checkout and benchmarking.

  10. Canadian CANDU fuel development programs and recent fuel operating experience

    This paper summarizes the performance and operating experience of CANDU fuel in Canadian CANDU reactors in 1999 and 2000. The extremely low rate of fuel defects continues to demonstrate that CANDU fuel is performing exceptionally well. Over the 2-year period, the fuel bundle defect rate for all bundles irradiated in Canadian CANDU reactors has remained very low, between 0.011% (suspected defects) and 0.007% (confirmed defects). On a fuel element basis, this represents a rate of confirmed defects of about 0.0002%; this rate is approaching 2 defects per million fuel elements! This successful performance is the result of a number of contributing factors, including a simple and robust fuel design with conservative design margins, reliable and specialized manufacturing processes that have been developed over the years, and fuel operations that conform to the fuel operating limits. Strong linkages between plant operation, designers, and Canadian fuel research and development programs also contribute to the high performance of the current CANDU fuel. The Fuel Design and Performance program, funded by the CANDU Owners Group, addresses licensing and operational issues that are common to the Canadian CANDU utilities. In addition, AECL's Fuel and Fuel Cycles working group directs R and D to support evolutionary improvements to the fuel products, as well as longer-term R and D for advanced fuel concepts. This paper describes the development programs in 1999/2000. (author)

  11. Canadian CANDU fuel development programs and recent fuel operating experience

    This paper summarizes the performance and operating experience of CANDU fuel in Canadian CANDU reactors in 1999 and 2000. The extremely low rate of fuel defects continues to demonstrate that CANDU fuel is performing exceptionally well. Over the 2-year period, the fuel bundle defect rate for all bundles irradiated in Canadian CANDU reactors has remained very low, between 0.011% (suspected defects) and 0.007% (confirmed defects). On a fuel element basis, this represents a rate of confirmed defects of about 0.0002%; this rate is approaching 2 defects per million fuel elements. This successful performance is the result of a number of contributing factors, including a simple and robust fuel design with conservative design margins, reliable and specialized manufacturing processes that have been developed over the years, and fuel operations that conform to the fuel operating limits. Strong linkages between plant operation, designers, and Canadian fuel research and development programs also contribute to the high performance of the current CANDU fuel. The Fuel Design and Performance program, funded by the CANDU Owners Group, addresses licensing and operational issues that are common to the Canadian CANDU utilities. In addition, AECL's Fuel and Fuel Cycles working group directs R and D to support evolutionary improvements to the fuel products, as well as long-term R and D for advanced fuel concepts. This paper describes the development programs in 1999/2000. (author)

  12. Preliminary Investigation on Turbulent Flow in Tight-lattice Rod Bundle with Twist-mixing Vane Spacer Grid

    Our research group has investigated the effect of P/D difference on the behavior of turbulent rod bundle flow without the mixing vane spacer grid, using PIV (Particle Image Velocimetry) and MIR (Matching Index of Refraction) techniques for tight lattice fuel rod bundle application. In this work, using the tight-lattice rod bundle with a twist-mixing vane spacer grid, the turbulent rod bundle flow is preliminarily examined to validate the PIV measurement and CFD (Computational Fluid Dynamics) simulation. The turbulent flow in the tight-lattice rod bundle with a twist-mixing vane spacer grid was preliminarily examined to validate the PIV measurement and CFD simulation. Both were in agreement with each other within a reasonable degree of accuracy. Using PIV measurement and CFD simulation tested in this work, the detailed investigations on the behavior of turbulent rod bundle flow with the twist-mixing vane spacer grid will be performed at various conditions, and reported in the near future

  13. Evaluation of existing correlations for the prediction of pressure drop in wire-wrapped hexagonal array pin bundles

    Highlights: • Wire-wrapped bundle friction factor data and correlations thoroughly collected. • Three methodologies proposed for identifying the best fit correlation. • 80 out of 141 bundles selected as database for evaluation. • The detailed Cheng and Todreas correlation identified to fit the data best. - Abstract: Existing wire-wrapped fuel bundle friction factor correlations were evaluated to identify their comparative fit to the available pressure drop experimental data. Five published correlations, those of Rehme (REH), Baxi and Dalle Donne (BDD, which used the correlations of Novendstern in the turbulent regime and Engel et al. in the laminar and transition regimes), detailed Cheng and Todreas (CTD), simplified Cheng and Todreas (CTS), and Kirillov (KIR, developed by Russian scientists) were studied. Other correlations applicable to a specific case were also evaluated but only for that case. Among all 132 available bundle data, an 80 bundle data set was judged to be appropriate for this evaluation. Three methodologies, i.e., the Prediction Error Distribution, Agreement Index and Credit Score were principally used for investigating the goodness of each correlation in fitting the data. Evaluations have been performed in two categories: 4 cases of general user interest and 3 cases of designer specific interest. The four general user interest cases analyzed bundle data sets in four flow regimes – i.e., all regimes, the transition and/or turbulent regimes, the turbulent regime, and the laminar regime. The three designer interest cases analyzed bundles in the fuel group, the blanket and control group and those with P/D > 1.06, for the transition/turbulent regimes. For all these cases, the detailed Cheng and Todreas correlation is identified as yielding the best fit. Specifically for the all flow regimes evaluation, the best fit correlation in descending order is CTD, BDD/CTS (tie), REH and KIR. For the combined transition/turbulent regime, the order is

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

    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

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

    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)

  16. Compactifications of reductive groups as moduli stacks of bundles

    Martens, Johan; Thaddeus, Michael

    Let G be a reductive group. We introduce the moduli problem of "bundle chains" parametrizing framed principal G-bundles on chains of lines. Any fan supported in a Weyl chamber determines a stability condition on bundle chains. Its moduli stack provides an equivariant toroidal compactification of ...

  17. VECTOR BUNDLE, KILLING VECTOR FIELD AND PONTRYAGIN NUMBERS

    周建伟

    1991-01-01

    Let E be a vector bundle over a compact Riemannian manifold M. We construct a natural metric on the bundle space E and discuss the relationship between the killing vector fields of E and M. Then we give a proof of the Bott-Baum-Cheeger Theorem for vector bundle E.

  18. Noncommutative principal torus bundles via parametrised strict deformation quantization

    Hannabuss, Keith; Mathai, Varghese

    2009-01-01

    In this paper, we initiate the study of a parametrised version of Rieffel's strict deformation quantization. We apply it to give a classification of noncommutative principal torus bundles, in terms of parametrised strict deformation quantization of ordinary principal torus bundles. The paper also contains a putative definition of noncommutative non-principal torus bundles.

  19. Geometry of torus bundles in integrable Hamiltonian systems

    Lukina, Olga

    2008-01-01

    Thesis is concerned with global properties of Lagrangian bundles, i.e. symplectic n-torus bundles, as these occur in integrable Hamiltonian systems. It treats obstructions to triviality and concerns with classification of such bundles, as well as with manifestations of global invariants in real-worl

  20. Experimental study for the effects of ballooned rod bundle on the convective heat transfer by single-phase steam flow

    For a large break loss-of-coolant accident (LBLOCA) conditions in a pressurized-water reactor, the cladding temperature increases until the reflood phase and the increased temperature can make a ballooned fuel rods. As a result, the flow passage area of sub-channel is reduced and it leads the redistribution of flow and heat transfer in sub-channels. During the single-phase steam flow in the early phase of the reflood, the cladding temperature may increase and have a peak value due to low heat transfer from the fuel to the steam. If a LBLOCA condition and ballooned fuel rods are occurred, the effect of reduced flow passage on the convective heat transfer by single-phase steam flow is important phenomena to analyze the safety of a reactor. The present experiments were performed in various Reynolds numbers (about 2600∼13000) to investigate the effect of the Ballooned fuel rods on heat transfer phenomena by single-phase steam flow. The experiments were performed in two rod bundles in KAERI reflood ATHER test facility. One is a non-deformed 6x6 rod bundle, which consists of 36 non-deformed heater rods. The other is a deformed 5x5 rod bundle that consists of 9 deformed heater rods and 16 non-deformed heater rods. The cladding temperature and convective heat transfer for two rod bundles are compared for each flow conditions and the effects of experimental parameters are analyzed. (author)

  1. Stability of Picard Bundle Over Moduli Space of Stable Vector Bundles of Rank Two Over a Curve

    Indranil Biswas; Tomás L Gómez

    2001-08-01

    Answering a question of [BV] it is proved that the Picard bundle on the moduli space of stable vector bundles of rank two, on a Riemann surface of genus at least three, with fixed determinant of odd degree is stable.

  2. The manufacturing role in fuel performance

    Manufacturing companies have been involved in the CANDU fuel industry for more than 40 years. Early manufacturing contributions were the development of materials and processes used to fabricate the CANDU fuel bundle. As CANDU reactors were commissioned, the manufacturing contribution has been to produce economical, high quality fuel for the CANDU market. (author)

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

    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

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

    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)

  5. Benchmark Experiments for Natural Convection in Nuclear Fuel Rod Bundles

    Jones, Kyle L.; Smith, Barton L

    2016-01-01

    Natural convection is a phenomenon in which a flow of the fluid surrounding a body is induced by a change in density due to the temperature difference between the body and the fluid. This flow can be highly non-linear and turbulent, generating eddies. The complex interaction between the convective, viscous and buoyant forces requires the use of modern turbulent simulation tools for simulation. The accuracy of these tools, due to non-linearity, is difficult to assess. The present study investi...

  6. Productivity and costs of slash bundling in Nordic conditions

    Kaerhae, K.; Vartiamaeki, T. [Metsaeteho Oy, P.O. Box 101, FI-00171 Helsinki (Finland)

    2006-12-15

    The number of slash bundlers and the volume of slash bundling have been rapidly increasing during the last few years in Finland. However, no comprehensive time or follow-up studies have been carried out on slash bundling technology in Finland or in any other country. Metsateho Oy carried out studies on the productivity and costs of slash bundling in different Nordic recovering conditions. The study methods included both time and follow-up studies. Data were collected during the summer and winter period primarily in Norway spruce (Picea abies L. Karst.) dominated clear cutting sites. The bundling techniques performed by different types of bundler (Fiberpac 370, Timberjack 1490D, Pika RS 2000, Valmet WoodPac) were studied. The average productivity of slash bundling was 18.1 bundles per operating (E{sub 15}, including delays shorter than 15min) hour with the Timberjack 1490D and Fiberpac 370 bundlers in the follow-up study. The operator of the slash bundler had the greatest effect on the productivity of bundling. The prerequisite for increased bundling volumes is a reduction in the costs of the most expensive sub-stage of the bundling supply chain, i.e. bundling itself. This requires improved recovery conditions at bundling sites, increased bundling productivity, larger sized bundles, and the execution of bundling operations in two work shifts using an efficient bundler and effective operator working methods. Implementation of these development measures will bring the bundling supply chain up to a speed that makes it the most competitive supply chain for forest chips in terms of total supply costs for long-distance transportation distances of more than 60km. (author)

  7. Bundling of harvesting residues and whole-trees and the treatment of bundles; Hakkuutaehteiden ja kokopuiden niputus ja nippujen kaesittely

    Kaipainen, H.; Seppaenen, V.; Rinne, S.

    1996-12-31

    The conditions on which the bundling of the harvesting residues from spruce regeneration fellings would become profitable were studied. The calculations showed that one of the most important features was sufficient compaction of the bundle, so that the portion of the wood in the unit volume of the bundle has to be more than 40 %. The tests showed that the timber grab loader of farm tractor was insufficient for production of dense bundles. The feeding and compression device of the prototype bundler was constructed in the research and with this device the required density was obtained.The rate of compaction of the dry spruce felling residues was about 40 % and that of the fresh residues was more than 50 %. The comparison between the bundles showed that the calorific value of the fresh bundle per unit volume was nearly 30 % higher than that of the dry bundle. This means that the treatment of the bundles should be done of fresh felling residues. Drying of the bundles succeeded well, and the crushing and chipping tests showed that the processing of the bundles at the plant is possible. The treatability of the bundles was also excellent. By using the prototype, developed in the research, it was possible to produce a bundle of the fresh spruce harvesting residues, the diameter of which was about 50 cm and the length about 3 m, and the rate of compaction over 50 %. By these values the reduction target of the costs is obtainable

  8. Nuclear fuel transporting container

    Purpose: To prevent the failure of nuclear fuel rods constituting a nuclear fuel assembly contained to the inside of a container upon fire accidents or the likes. Constitution: The nuclear fuel transportation container comprises a tightly sealed inner vessel made of steels for containing a nuclear fuel assembly consisting of bundled nuclear fuel rods, a heat shielding material surrounding the inner vessel, shock absorber and an outer vessel. A relief safety valve is disposed to the inner vessel that actuates at a specific pressure higher than the normal inner pressure for the nuclear fuel rods of the fuel assembly and lower than the allowable inner pressure of the inner vessel. The inside of the inner vessel is pressurized by way of the safety valve such that the normal inner pressure in the inner vessel is substantially equal to the normal inner pressure for the nuclear fuel rods. (Aizawa, K.)

  9. Defective fuel location by dry sipping

    One of the largest fuel defect excursions experienced by Ontario Hydro occurred in Unit 3 of Bruce NGS-A during December 1983 and early 1984. A large part of our response to this problem involved the underwater visual examinations of irradiated fuel bundles. We were able to greatly reduce our workload by monitoring the radioactivity released by defected fuel bundles during their dry transfer from the D2O of the fueling machine to the H2O of the irradiated fuel bay (IFB). This procedure is commonly called 'dry sipping'. We were very successful in correlating dry sipping indications from a suspect bundle pair to visually confirmed defected bundles. Residual contamination in the active vent of the fuel transfer system, after discharge of fuel to the IFB, almost always indicated discharge of a defected bundle. This was a new observation. The short lived activity spike in the vent, during the discharge, was not found to be as reliable an indicator. This successful experience contributed significantly to the resolution of the fuel defect problem. Also, the automatic system we installed greatly reduced the radiological hazard to station staff from the manual dry sipping procedure that existed previously. These results promise useful applications to other Ontario Hydro nuclear stations

  10. Comparison of ASSERT subchannel code with Marviken bundle data

    In this paper ASSERT predictions are compared with the Marviken 6-rod bundle and 36+1 rod bundle. The predictions are presented for two experiments in the 6-rod bundle and four experiments in the 36+1 rod bundle. For low inlet subcooling, the void predictions are in good agreement with the experimental data. For high inlet subcooling, however, the agreement is not as good. This is attributed to the fact that in the high inlet subcooling experiments, single phase turbulent mixing plays a more important role in determining flow conditions in the bundle

  11. Multiwalled carbon nanotube reinforced biomimetic bundled gel fibres.

    Kim, Young-Jin; Yamamoto, Seiichiro; Takahashi, Haruko; Sasaki, Naruo; Matsunaga, Yukiko T

    2016-08-19

    This work describes the fabrication and characterization of hydroxypropyl cellulose (HPC)-based biomimetic bundled gel fibres. The bundled gel fibres were reinforced with multiwalled carbon nanotubes (MWCNTs). A phase-separated aqueous solution with MWCNT and HPC was transformed into a bundled fibrous structure after being injected into a co-flow microfluidic device and applying the sheath flow. The resulting MWCNT-bundled gel fibres consist of multiple parallel microfibres. The mechanical and electrical properties of MWCNT-bundled gel fibres were improved and their potential for tissue engineering applications as a cell scaffold was demonstrated. PMID:27200527

  12. Effectiveness of Hair Bundle Motility as the Cochlear Amplifier

    Sul, Bora; Iwasa, Kuni H.

    2009-01-01

    The effectiveness of hair bundle motility in mammalian and avian ears is studied by examining energy balance for a small sinusoidal displacement of the hair bundle. The condition that the energy generated by a hair bundle must be greater than energy loss due to the shear in the subtectorial gap per hair bundle leads to a limiting frequency that can be supported by hair-bundle motility. Limiting frequencies are obtained for two motile mechanisms for fast adaptation, the channel re-closure mode...

  13. Anatomic Double-Bundle Posterior Cruciate Ligament Reconstruction.

    Chahla, Jorge; Nitri, Marco; Civitarese, David; Dean, Chase S; Moulton, Samuel G; LaPrade, Robert F

    2016-02-01

    The posterior cruciate ligament (PCL) is known to be the main posterior stabilizer of the knee. Anatomic single-bundle PCL reconstruction, focusing on reconstruction of the larger anterolateral bundle, is the most commonly performed procedure. Because of the residual posterior and rotational tibial instability after the single-bundle procedure and the inability to restore the normal knee kinematics, an anatomic double-bundle PCL reconstruction has been proposed in an effort to re-create the native PCL footprint more closely and to restore normal knee kinematics. We detail our technique for an anatomic double-bundle PCL reconstruction using Achilles and anterior tibialis tendon allografts. PMID:27284530

  14. Critical heat flux and pressure drop for a CANFLEX bundle string inside an axially non-uniform flow channel

    Experimental data of dryout power and pressure drop have been obtained with a simulated string of twelve aligned, full-scale, CANFLEX fuel bundles. The bundle string consisted of 43 elements and was equipped with junction and appendages simulations. It was installed inside three flow tubes simulating three different creep profiles: one had a uniform inside diameter of 103.86 mm and the other two had axially varying inside diameters, with a peak of either 107.29 mm or 109.16 mm (3.3% and 5.1% larger than the uniform tube). Pressure variations along the fuel string were obtained with differential-pressure cells connected to a number of pressure taps. Sliding thermocouples were used to obtain surface-temperature measurements and detect dryout. A wide range of steam-water flow conditions was covered in the current tests: an outlet-pressure range from 6 to 11 MPa, a mass-flow-rate range from 7 to 25 kg/s, and an inlet-fluid-temperature range from 200 to 290 deg C. This paper focuses primarily on data obtained at normal operating pressures with the axially non-uniform channel that had a maximum diameter 5.1% larger than the reference pressure tube. Local and boiling-length-average (BLA) critical-heat-flux values were derived from the dryout-power data for various flow conditions. Unlike the traditional BLA approach, the averaging process was initiated from the onset of significant void (OSV), instead of from the saturation point. This allowed the extension of the BLA approach to subcooled dryout conditions. The OSV values were evaluated from the pressure distribution along the bundle string. Comparisons of various parameters were made between the 37-element and CANFLEX bundles. Overall, the dryout-power values were consistently higher for the CANFLEX bundle than the 37-element bundle. At inlet-flow conditions of interest, the dryout-power measurements were, on average, 17% higher for the CANFLEX bundle than the 37-element bundle. The fuel-string pressure drop was similar

  15. The turbulent flow in rod bundles

    Experimental studies have shown that the axial and azimuthal turbulence intensities in the gap regions of rod bundles increase strongly with decreasing rod spacing; the fluctuating velocities in the axial and azimuthal directions have a quasi-periodic behaviour. To determine the origin of this phenomenon, an its characteristics as a function of the geometry and the Reynolds number, an experimental investigation was performed on the turbulent in several rod bundles with different aspect ratios (P/D, W/D). Hot-wires and microsphones were used for the measurements of velocity and wall pressure fluctuations. The data were evaluated to obtain spectra as well as auto and cross correlations. Based on the results, a phenomenological model is presented to explain this phenomenon. By means of the model, the mass exchange between neighbouring subchannels is explained

  16. Reactor application of an improved bundle divertor

    A Bundle Divertor was chosen as the impurity control and plasma exhaust system for the beam driven Demonstration Tokamak Hybrid Reactor - DTHR. In the context of a preconceptual design study of the reactor and associated facility a bundle divertor concept was developed and integrated into the reactor system. The overall system was found feasible and scalable for reactors with intermediate torodial field strengths on axis. The important design characteristics are: the overall average current density of the divertor coils is 0.73 kA for each tesla of toroidal field on axis; the divertor windings are made from super-conducting cables supported by steel structures and are designed to be maintainable; the particle collection assembly and auxiliary cryosorption vacuum pump are dual systems designed such that they can be reactivated alterntively to allow for continuous reactor operation; and the power requirement for energizing and operating the divertor is about 5 MW

  17. Venereau polynomials and related fiber bundles

    Kaliman, Shulim; ZAIDENBERG, MIKHAIL

    2003-01-01

    The Venereau polynomials v-n:=y+x^n(xz+y(yu+z^2)), n>= 1, on A4 have all fibers isomorphic to the affine space A3. Moreover, for all n>= 1 the map (v-n, x) : A4 -> A2 yields a flat family of affine planes over A2. In the present note we show that over the punctured plane A2\\0, this family is a fiber bundle. This bundle is trivial if and only if v-n is a variable of the ring C[x][y,z,u] over C[x]. It is an open question whether v1 and v2 are variables of the polynomial ring C[x,y,z,u]. S. Vene...

  18. A fibre bundle formulation of quantum geometry

    Quantum geometries whose points are stochastic and serve as seats for quantum space-time excitons are formulated as fibre bundles over base spaces of mean values with a Minkowski or general relativistic structure. The fibres contain the proper wave functions of all exciton states in a given model. The notion of covariance and propagation in quantum space-times constituting such fibre bundles is investigated. Maxwell and Yang-Mills gauge degrees of freedom are introduced by appropriately enlarging the structure group, which in all cases contains phase-space representations of the Poincare group corresponding to the exciton wave function sample space specific to a given model. It is shown that these formulations give rise in a natural manner to certain realizations of the relativistic canonical commutation relations in terms of covariant derivatives involving internal as well as external degrees of freedom of space-time excitons

  19. Heterotic String Compactification and New Vector Bundles

    Lin, Hai; Wu, Baosen; Yau, Shing-Tung

    2016-07-01

    We propose a construction of Kähler and non-Kähler Calabi-Yau manifolds by branched double covers of twistor spaces. In this construction we use the twistor spaces of four-manifolds with self-dual conformal structures, with the examples of connected sum of n {mathbb{P}2}s. We also construct K3-fibered Calabi-Yau manifolds from the branched double covers of the blow-ups of the twistor spaces. These manifolds can be used in heterotic string compactifications to four dimensions. We also construct stable and polystable vector bundles. Some classes of these vector bundles can give rise to supersymmetric grand unified models with three generations of quarks and leptons in four dimensions.

  20. Rod bundle burnout data and correlation comparisons

    Rod bundle burnout data from 30 steady-state and 3 transient tests were obtained from experiments performed in the Thermal Hydraulic Test Facility at the Oak Ridge National Laboratory. The tests covered a parameter range relevant to intact core reactor accidents ranging from large break to small break loss-ofcoolant conditions. Instrumentation within the 64-rod test section indicated that burnout occurred over an axial range within the bundle. The distance from the point where the first dry rod was detected to the point where all rods were dry was up to 60 cm in some of the tests. The burnout data should prove useful in developing new correlations for use in reactor thermalhydraulic codes. Evaluation of several existing critical heat flux correlations using the data show that three correlations, the Barnett, Bowring, and Katto correlations, perform similarly and correlate the data better than the Biasi correlation

  1. Client Provider Collaboration for Service Bundling

    LETIA, I. A.

    2008-04-01

    Full Text Available The key requirement for a service industry organization to reach competitive advantages through product diversification is the existence of a well defined method for building service bundles. Based on the idea that the quality of a service or its value is given by the difference between expectations and perceptions, we draw the main components of a frame that aims to support the client and the provider agent in an active collaboration meant to co-create service bundles. Following e3-value model, we structure the supporting knowledge around the relation between needs and satisfying services. We deal with different perspectives about quality through an ontological extension of Value Based Argumentation. The dialog between the client and the provider takes the form of a persuasion whose dynamic object is the current best configuration. Our approach for building service packages is a demand driven approach, allowing progressive disclosure of private knowledge.

  2. Radiological evidence for the triple bundle anterior cruciate ligament.

    MacKay, James W; Whitehead, Harry; Toms, Andoni P

    2014-10-01

    The anterior cruciate ligament (ACL) has traditionally been described as having two bundles--one anteromedial and one posterolateral. This has been challenged by studies proposing the existence of a third, intermediate, bundle with distinct functional significance, an arrangement that has been described in a number of domesticated animal species. No radiological evidence for the triple bundle ACL has previously been described. A prevalence study was carried out on 73 consecutive human knee magnetic resonance (MR) studies to determine the number of visible bundles, excluding individuals with a history of ACL injury or mucoid degeneration. A triple bundle ACL was demonstrated in 15 out of 73 human knees (20.5%, 95% confidence interval 12.9-31.2%). This is the first radiological description of the human triple bundle ACL. There was MR imaging evidence of a triple bundle ACL in approximately one fifth of human knees in this study. PMID:24890455

  3. Critical heat flux measurements in rod bundles using light water and heavy water as coolant

    A series of Critical Heat Flux (CHF), subchannel mixing and pressure drop tests were performed on a full scale simulated 37 element fuel bundle at the Heat Transfer Research Facility of Columbia University for Siemens/Kraftwerk Union Ag. of Germany. The experimental program consisted of thermal hydraulic testing of a full scale 37 element fuel bundle 3000 mm long to determine two phase pressure drop and CHF characteristics. The test were performed on the simulated fuel bundle in a vertical circular test housing with an outer diameter of 108.25 mm. The axial power distribution was uniform, while the radial power distribution was non-uniform with a heat flux depression of about 35%. The bundle geometry was maintained by spacer grids. In this experimental program, 92 CHF runs were performed using light water (H2O) as the coolant, and 22 CHF runs were performed using heavy water (D2O) as the coolant to reduce the risk of heavy water loss. The CHF tests covered the following parameter ranges: pressure from 70 to 150 bars in light water and at 100 and 115 bars in heavy water; mass velocities from 1000 to 5600 kg/m2s, and inlet temperatures from 200 to 320 degree C. The experiments carried out in heavy water were in essence a duplication of the test conditions of the ones obtained in light water, facilitating a direct comparison of the results using both fluids. Furthermore, an evaluation of the experiments using the subchannel analysis method was made with the results being presented. Analysis shows that CHF with heavy water as a coolant can be predicted with sufficient accuracy by applying Ahmad's scaling laws for fluid -to fluid modeling. Therefore, CHF correlations developed and verified for light water reactors can be applied to design heavy water cooled reactor cores

  4. CHF and flow instability in rod bundles

    Data for two very different rod bundles have been analyzed using a new CHF correlation and a crude, but simple, subchannel analysis. The CHF correlation was developed for round uniform tubes and has been shown to accurately predict CHF in nonuniform tubes. The first set of data was for a KWU rod bundle (37 rods) with a heated length of 3.00 m and an O.D. (outside diameter) of 12.9 mm over a range of pressure 70 to 150 bar in upflow. The second set of data was for a 5 x 5 TRIGA rod bundle with a heated length of 0.559 m and 13.75 mm O.D. over a range of pressure of 0.945 to 1.372 bar in downflow. In contrast to the KWU data, the correlation greatly over estimates the CHF values for the TRIGA data. The TRIGA CHF data correlate very well with the variable qsat assuming no mixing, qc,exp = 0.955qsat (stdev = 9.87%). This result strongly suggests that these instabilities, which resulted immediately in CHF, are triggered by the Onset of Flow Instability (OFI) rather than CHF. The wide spread in rod power factors, the low pressure, and the downflow condition all contribute to promoting this type of instability (Ledinegg). The crude subchannel analysis has been compared with calculations of exit conditions of the hot channel using COBRA code. The agreement is fair when the homogeneous equilibrium model is used in the COBRA code. This is expected since the exit of the hot channel is always subcooled. Using Zuber's, along with other, void fraction relations in COBRA yields much lower exit velocities and high positive exit qualities, and, in some cases, convergence difficulties arise. The facts indicate that the bundle has already past the OFI point: which is possible since no CHF calculation was made in these COBRA analyses. (J.P.N)

  5. Interstitial He and Ne in Nanotube Bundles

    Stan, G.; Crespi, V. H.; Cole, M. W.; Boninsegni, M.

    1998-01-01

    We explore the properties of atoms confined to the interstitial regions within a carbon nanotube bundle. We find that He and Ne atoms are of ideal size for physisorption interactions, so that their binding energies are much greater there than on planar surfaces of any known material. Hence high density phases exist at even small vapor pressure. There can result extraordinary anisotropic liquids or crystalline phases, depending on the magnitude of the corrugation within the interstitial channels.

  6. Effective freeness of adjoint line bundles

    Heier, Gordon

    2001-01-01

    In this note we establish a new Fujita-type effective bound for the base point freeness of adjoint line bundles on a compact complex projective manifold of complex dimension $n$. The bound we obtain (approximately) differs from the linear bound conjectured by Fujita only by a factor of the cube root of $n$. As an application, a new effective statement for pluricanonical embeddings is derived.

  7. On Complex Supermanifolds with Trivial Canonical Bundle

    Groeger, Josua

    2016-01-01

    We give an algebraic characterisation for the triviality of the canonical bundle of a complex supermanifold in terms of a certain Batalin-Vilkovisky superalgebra structure. As an application, we study the Calabi-Yau case, in which an explicit formula in terms of the Levi-Civita connection is achieved. Our methods include the use of complex integral forms and the recently developed theory of superholonomy.

  8. Imperfect Bundling In Public-Private Partnerships

    Luciano Greco

    2012-01-01

    The economic literature on PPPs has generally overlooked agency problems within private consortia. We provide a first contribution in this direction, relying on a simple incomplete contracts framework where a Builder and an Operator set up a Special Purpose Vehicle (SPV) to carry out a contract with the government. Because of incomplete contracts, the bundling of tasks is imperfect, and the SPV ownership structure is the main tool to regulate the power of private incentives. The scope for wel...

  9. Uncovering ecosystem service bundles through social preferences.

    Berta Martín-López

    Full Text Available Ecosystem service assessments have increasingly been used to support environmental management policies, mainly based on biophysical and economic indicators. However, few studies have coped with the social-cultural dimension of ecosystem services, despite being considered a research priority. We examined how ecosystem service bundles and trade-offs emerge from diverging social preferences toward ecosystem services delivered by various types of ecosystems in Spain. We conducted 3,379 direct face-to-face questionnaires in eight different case study sites from 2007 to 2011. Overall, 90.5% of the sampled population recognized the ecosystem's capacity to deliver services. Formal studies, environmental behavior, and gender variables influenced the probability of people recognizing the ecosystem's capacity to provide services. The ecosystem services most frequently perceived by people were regulating services; of those, air purification held the greatest importance. However, statistical analysis showed that socio-cultural factors and the conservation management strategy of ecosystems (i.e., National Park, Natural Park, or a non-protected area have an effect on social preferences toward ecosystem services. Ecosystem service trade-offs and bundles were identified by analyzing social preferences through multivariate analysis (redundancy analysis and hierarchical cluster analysis. We found a clear trade-off among provisioning services (and recreational hunting versus regulating services and almost all cultural services. We identified three ecosystem service bundles associated with the conservation management strategy and the rural-urban gradient. We conclude that socio-cultural preferences toward ecosystem services can serve as a tool to identify relevant services for people, the factors underlying these social preferences, and emerging ecosystem service bundles and trade-offs.

  10. Noncommutative line bundle and Morita equivalence

    Jurco, Branislav; Schupp, Peter; Wess, Julius

    2001-01-01

    Global properties of abelian noncommutative gauge theories based on $\\star$-products which are deformation quantizations of arbitrary Poisson structures are studied. The consistency condition for finite noncommutative gauge transformations and its explicit solution in the abelian case are given. It is shown that the local existence of invertible covariantizing maps (which are closely related to the Seiberg-Witten map) leads naturally to the notion of a noncommutative line bundle with noncommu...

  11. Bundling harvester; Harvennuspuun automaattisen nippukorjausharvesterin kehittaeminen

    Koponen, K. [Eko-Log Oy, Kuopio (Finland)

    1997-12-01

    The starting point of the project was to design and construct, by taking the silvicultural point of view into account, a harvesting and processing system especially for energy-wood, containing manually driven bundling harvester, automating of the harvester, and automated loading. The equipment forms an ideal method for entrepreneur`s-line harvesting. The target is to apply the system also for owner`s-line harvesting. The profitability of the system promotes the utilisation of the system in both cases. The objectives of the project were: to construct a test equipment and prototypes for all the project stages, to carry out terrain and strain tests in order to examine the usability and durability, as well as the capacity of the machine, to test the applicability of the Eko-Log system in simultaneous harvesting of energy and pulp woods, and to start the marketing and manufacturing of the products. The basic problems of the construction of the bundling harvester have been solved using terrain-tests. The prototype machine has been shown to be operable. Loading of the bundles to form sufficiently economically transportable loads has been studied, and simultaneously, the branch-biomass has been tried to be utilised without loosing the profitability of transportation. The results have been promising, and will promote the profitable utilisation of wood-energy. (orig.)

  12. Nuclear reactor control bundle guide system

    Each bundle is formed by several absorbent rods, which are vertically movable and are connected together by a spider to a common axial operating rod, and guide means for the control bundles in their displacement, out of the core; the said means comprise guide boxes containing horizontal plates for discontinuous guiding, at the upper part of the boxes, of absorbent rods positioned in pairs on a radius and individual peripheral absorbent rods of the control bundle. At the lower part of the boxes in a continuous guiding zone, guiding of the absorbent rods positioned in pairs on a radius is effected by association of the horizontal plates for mechanical guiding of the rods, with housings which minimise hydraulic effects by smoothing the coolant flow in the radial direction around the absorbent rods. The hydraulic housings are mounted between the horizontal plates as discontinuous spacers. Pressure differences around each rod are minimised or eliminated and continuous guiding is achieved without affecting the design of the guide boxes, the internal equipment or the pressure vessel. The invention can be applied to PWRs

  13. Fuel assembly

    The object of the present invention is to improve the hydrodynamic stability in the fuel channels of BWR type reactors and effectively utilize the coolant driving power corresponding to the reduction due to pressure loss. That is, in a fuel assembly having usual fuel rods and, in addition, water rods and short fuel rods, the structures of water rods, upper tie plates and the spacers are designed from a hydrodynamic point of view, to reduce the pressure loss. On the other hand, a lattice-like flow channel resistance member is disposed to a lower tie plate. The bundle flow rate is made uniform by the flow channel resistance member, and the pressure loss of the tie plate is increased by the reduction of the pressure loss by the arrangement of the short fuel rod and the reduction of the pressure loss described above. Since this increases the ratio of the single phase stream pressure loss in the total reactor core pressure loss, the hydrodynamic stability in the fuel channel is improved. (I.J.)

  14. Semi-empirical model for the calculation of flow friction factors in wire-wrapped rod bundles

    LMFBR fuel elements consist of wire-wrapped rod bundles, with triangular array, with the fluid flowing parallel to the rods. A semi-empirical model is developed in order to obtain the average bundle friction factor, as well as the friction factor for each subchannel. The model also calculates the flow distribution factors. The results are compared to experimental data for geometrical parameters in the range: P(div)D = 1.063 - 1.417, H(div)D = 4 - 50, and are considered satisfactory. (Author)

  15. Thermal-hydraulic analysis of three-dimensional natural convection in a vertical rod bundle

    Numerical analyses have been conducted in boundary-fitted coordinates on the problem of three-dimensional natural convection in a vertical rod bundle that consists of a group of seven heated rods and a cold hexagonal enclosure surrounding them. Flow patterns and temperature distributions are obtained for Rayleigh numbers up to 107 with the aspect ratio of vertical to horizontal dimensions being set at the value of unity. Three Rayleigh number regions of different modes of heat transfer, i.e., pseudo-conduction, transitional convection and boundary layer regions, are discussed. The flow structures and the effect of natural convection on local and overall heat transfer rates are presented. It is found that the rods each tend to assume nearly the same average heat flux rate (in the case of isothermal conditions) as the Rayleigh number is increased. The results obtained enhance the basic understanding of buoyancy-induced fluid flow and heat transfer in rod bundles, and in particular the differences that are obtained when different boundary conditions are imposed on the inner cylindrical rods. Natural convection in an enclosed space consisting of a hot vertical rod bundle placed in a cold enclosure has potential applications including storage of heat-generating spent fuel assemblies, and heat removal of a nuclear fuel-pin configuration in a light-water reactor in situations of emergency

  16. Combustor and method for distributing fuel in the combustor

    Uhm, Jong Ho; Ziminsky, Willy Steve; Johnson, Thomas Edward; York, William David

    2016-04-26

    A combustor includes a tube bundle that extends radially across at least a portion of the combustor. The tube bundle includes an upstream surface axially separated from a downstream surface. A plurality of tubes extends from the upstream surface through the downstream surface, and each tube provides fluid communication through the tube bundle. A baffle extends axially inside the tube bundle between adjacent tubes. A method for distributing fuel in a combustor includes flowing a fuel into a fuel plenum defined at least in part by an upstream surface, a downstream surface, a shroud, and a plurality of tubes that extend from the upstream surface to the downstream surface. The method further includes impinging the fuel against a baffle that extends axially inside the fuel plenum between adjacent tubes.

  17. Heat transfer in a vertical 7-element bundle cooled with supercritical Freon-12

    Currently, SuperCritical Water-cooled nuclear Reactor (SCWR) concepts are being developed worldwide with an objective to increase thermal efficiencies of future Nuclear Power Plants (NPPs) on 10 -15% compared to those of current water-cooled NPPs. With such an increase in the thermal efficiencies, SCW NPPs will be at the current level of the most advanced thermal power plants: coal-fired SCW NPPs and combined-cycle gas-fired NPPs. However, to be able to develop SCWRs at least several key technical problems should be resolved. One of these problems is limited amount of experimental data on heat transfer in fuel bundles and based on that SCW heat-transfer correlations. Experiments in SCW are very complicated and expensive due to high critical parameters of water (pressure 22.064 MPa and temperature 374.95°C). Moreover, there are only few SCW test rigs, which capable to perform experiments in full-scale bundles. As a preliminary approach supercritical-pressure heat-transfer experiments in bundles can be performed in modeling fluids such as Freons or carbon dioxide. Therefore, a set of experimental data was obtained in Freon-12-cooled bundle simulator at the Institute of Physics and Power Engineering (IPPE, Obninsk, Russia). A vertical 7-element bundle was installed in a hexagonal flow channel. The test section consisted of elements that were 9.5 mm in diameter with the total heated length of 1 m. Bulk-fluid and wall temperature profiles were recorded using thermocouples. Several heat-transfer regimes were tested. Also, this paper references thermophysical properties of supercritical Freon-12 at the critical pressure (4.1361 MPa) and test pressure of 4.65 MPa. (author)

  18. Experimental benchmark data for PWR rod bundle with spacer-grids

    In numerical simulations of fuel rod bundle flow fields, the unsteady Navier–Stokes equations have to be solved in order to determine the time (phase) dependent characteristics of the flow. In order to validate the simulations results, detailed comparison with experimental data must be done. Experiments investigating complex flows in rod bundles with spacer grids that have mixing devices (such as flow mixing vanes) have mostly been performed using single-point measurements. In order to obtain more details and insight on the discrepancies between experimental and numerical data as well as to obtain a global understanding of the causes of these discrepancies, comparisons of the distributions of complete phase-averaged velocity and turbulence fields for various locations near spacer-grids should be performed. The experimental technique Particle Image Velocimetry (PIV) is capable of providing such benchmark data. This paper describes an experimental database obtained using two-dimensional Time Resolved Particle Image Velocimetry (TR-PIV) measurements within a 5 × 5 PWR rod bundle with spacer-grids that have flow mixing vanes. One of the unique characteristic of this set-up is the use of the Matched Index of Refraction technique employed in this investigation to allow complete optical access to the rod bundle. This unique feature allows flow visualization and measurement within the bundle without rod obstruction. This approach also allows the use of high temporal and spatial non-intrusive dynamic measurement techniques namely TR-PIV to investigate the flow evolution below and immediately above the spacer. The experimental data presented in this paper includes explanation of the various cases tested such as test rig dimensions, measurement zones, the test equipment and the boundary conditions in order to provide appropriate data for comparison with Computational Fluid Dynamics (CFD) simulations. Turbulence parameters of the obtained data are presented in order to gain

  19. Adsorption of Argon on Carbon nanotube bundles and its influence on the bundle lattice parameter

    We report experimental studies of the adsorption characteristics and structure of both Ar36 and Ar40 on single-wall carbon nanotube bundles. The structural studies make use of the large difference in coherent neutron scattering cross section for the two Ar isotopes to explore the influence of the adsorbate on the nanotube lattice parameter. We observe no dilation of the nanotube lattice with Ar40, and explain the apparent expansion of this lattice upon Ar36 adsorption by the location of the adsorbed Ar atoms on the outer bundle surface

  20. Modelling nuclear fuel vibrations in horizontal CANDU reactors

    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)