WorldWideScience

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. 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)

  15. 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)

  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. 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)

  20. 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)

  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. 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)

  13. 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.

  14. 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

  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. 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)

  12. 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

  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

    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.

  3. 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

  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. 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)

  14. 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

  15. 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)

  16. 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)

  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. 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)

  4. 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

  5. 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)

  6. 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)

  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. 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.)

  14. 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)

  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

    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

    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

  5. 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

  6. 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

  7. 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

  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. 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)

  17. 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)

  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. 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)

  5. 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)

  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. 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 ...

  17. 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)

  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. 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

  13. 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)

  14. 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)

  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

    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

  17. 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

  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. 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)

  8. 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)

  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)