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Sample records for candu fuel channel

  1. Design and Development of a Robotic Crawler for CANDU Fuel Channel Inspection

    Science.gov (United States)

    Shukla, Shivam

    For the design of a new robotic crawler drive unit for CANDU fuel channel inspection, a complete design and screening process was done in order to fulfil the objective of this research. A brief explanation of CANDU reactors is provided along with a discussion of the inspection systems that are currently in use. A study of some existing inspection systems is presented which was used for the development of the new robotic crawler design. A number of concepts were generated which underwent a screening process with the help of two design tools. With the help of these tools, a concept was chosen as the final design and details of it are presented. To demonstrate a proof-of-concept, the physical prototype of the robotic crawler was manufactured and assembled. A speed controller was implemented in the final design of the robotic crawler. A set of test procedures were performed on the final design and the results are discussed. Some improvements that can be done on the final design of the robotic crawler are also discussed in the final section of this thesis.

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

    Directory of Open Access Journals (Sweden)

    Doddy Kastanya

    2017-02-01

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

  3. Dimensional Measurements of Fresh CANDU Fuel Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Ji Su; Jo, Chang Keun; Jung, Jong Yeob; Koo, Dae Seo; Cho, Moon Sung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    This paper intends to provide the dimensional measurements of fresh CANDU fuel (37-element) bundle for the estimation of deformation of post-irradiated (PI) bundle. It is expensive and difficult to measure the fretting wear of bearing pad, the element bowing and the waviness of endplate at the two-phase high flow condition (above 24 kg/s) of out-of-reactor test. So, it is recommended to compare the geometry of fresh bundle with that of PI bundle to estimate the integrity of fuel bundle in the CANDU-6 fuel channel with two-phase flow condition. The measurement system has been developed to provide the visual inspection and the dimensional measurements within the accuracy of 10 {mu}m. It is applicable in-air and underwater to the CANDU bundle as well as the CANFLEX bundle. The in-air measurements of the 36 fresh CANDU bundles (S/N: B400892 {approx} B400927) are done by this system from February 2004 to March 2004 in the PHWR fresh fuel storage building of KNFC. These bundles are produced by KNFC manufacturing procedure and are waiting for the delivery to the Wolsong-3 plant, and are planned to load into the proposed test channels. The detail measurements contain the outer rod profile (including the bearing pad), the diameter of bundle, the bowing of bundle, the rod length and the surface profile of end plate (waviness)

  4. Fuel condition in Canadian CANDU 6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hu, R.H.; Macici, N [Hydro-Quebec, Montreal, Quebec (Canada); Gibb, R. [New Brunswick Power, Lepreau, NB (Canada); Purdy, P.L.; Manzer, A.M. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Kohn, E. [Ontario Hydro, Toronto, Ontario (Canada)

    1997-07-01

    The cornerstone of the CANDU concept is its natural uranium fuel, and the success of its reactor operation hinges on the fuel condition in the reactor. Neutron economy, on power refuelling, and simple fuel design are among the unique characteristics of CANDU fuel. In Canadian CANDU 6 reactors (Gentilly 2 and Point Lepreau), the 37-element fuel has provided an enviable record of safe, economic and reliable plant operation for 29 reactor years to date. The fuelling cost is among the lowest in the world - a corollary of high neutron economy, simple fuel design, and judicial fuelling scheme. The reliability of fuel is high: only 21 of the 60000 bundles discharged from Gentilly 2 were confirmed defective and the five-year period from March 1992 to February 1997 saw no defect at all at Gentilly-2. Also, thanks to the inherent on-power refuelling capability and an effective defect detection and removal system, the primary coolant loops are kept extremely clean (very low activity level) - benefiting both maintenance and safety. Moreover, the inventories of fission products in the core and in the channel are maintained within the safety analysis envelope, due to on-power fuelling and sophisticated fuel management. In this paper, CANDU 6 fuel performance is reviewed against the feedback from post-irradiation examinations, and the findings from our ongoing R and D program. The results suggest that the fuel behavior m reactor are basically as originally anticipated, despite an evolutionary 3% increase in bundle uranium mass in the 1980's. For operating conditions within the CANDU 6 37-element experience, the average strains are typically 0.09%; and fission gas release, 2.7%. The UO{sub 2} fuel remains stoichiometric after irradiation. In-core measurements of pressure tube fitting are generally low. All these observations are consistent with the excellent fuel performance statistics coming out of the two Canadian CANDU 6 reactors. Additionally, this paper will briefly

  5. Verification tests for CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  6. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  7. Fuel management simulation for CANFLEX-RU in CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  8. Experimental and numerical investigations of three-dimensional turbulent flow of water surrounding a CANDU simulation fuel bundle structure inside a channel

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, F.; Yu, S.D. [Department of Mechanical and Industrial Engineering, Ryerson University, 350 Victoria Street, Toronto, Ontario, M5B 2K3 (Canada); Cao, J. [Department of Mechanical and Industrial Engineering, Ryerson University, 350 Victoria Street, Toronto, Ontario, M5B 2K3 (Canada)], E-mail: jcao@ryerson.ca

    2009-11-15

    Computational fluid dynamics (CFD) is used to simulate highly turbulent coolant flows surrounding a simulation CANDU fuel bundle structure inside a flow channel. Three CFD methods are used: large eddy simulation (LES), detached eddy simulation (DES), and Reynolds stress model (RSM). The outcome of the simulations is compared with the experimental pressure data measured using an in-water microphone and a miniature pressure transducer placed at various locations in the vicinity of the bundle structure. Among all the three methods employed in developing computational models, LES provides the most accurate results for turbulent pressures.

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

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2015-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  11. Diagnostic technology for degradation of feeder pipes and fuel channels in CANDU reactor; development of aging assessment technology for CANDU pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Kim, Yun Jae; Huh, Nam Su; Kwak, Sang Log; Lee, Kyu Ho [Sungkyunkwan University, Seoul (Korea)

    2002-04-01

    This research project attempts to resolve two issues related to integrity assessment of CANDU pressure tubes; (1) FE analysis of blister formation and growth, and (2) engineering estimation scheme to predict creep deflection of pressure tubes. Results for blister formation and growth can be summarised as follows. Comparing the results from the FE analysis, developed within this project, with experimental data shows some differences ranging from 10-57%. Such difference results from two possible sources. One source is neglecting two phase diffusion. The present FE analysis considers only single phase diffusion, and thus blister growth can not be accurately modeled. The other source would be inherent errors associated with experimental measurement. Thus it has been concluded that further efforts should be made on two phase diffusion modeling. For developing mechanistic model of creep deflection, the proposed reference stress based model is simple to use. Extensive validation against creep FE results shows that the proposed model is also quite accurate. More important aspect of the proposed method is that it can be easily generalized to more complex problems. Thus it is believed that the present results provide a sound basis for sagging assessment of CANDU pressure tubes. 16 refs., 12 figs., 6 tabs. (Author)

  12. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR. PART 9 - CUTTING AND EXTRACTING DEVICE FUNCTIONING

    Directory of Open Access Journals (Sweden)

    Constantin POPESCU

    2015-05-01

    Full Text Available This paper presents a constructive solution proposed by the authors in order to achieve of a cutting and extracting device for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The Cutting and Extraction Device (CED performs the dismantling, cutting and extraction of the fuel channel components. It's a flexible and modular device, which is designed to work inside the fuel channel and has the following functions: moving with variable speed, temperature monitoring and video surveillance inside the pipe, unblock and extract the channel closure plug (from End Fitting - EF, unblock and extract the channel shield plug (from Lattice Tube - LT, block and cut the middle of the pressure tube, block and cut the end of the pressure tube, block and extract the half of pressure tube. The Cutting and Extraction Device (CED consists of following modules: guiding-fixing module, traction modules, cutting module, guiding-extracting module and flexible elements for modules connecting. The guiding-fixing module is equipped with elastic guiding rollers and fixing claws in working position, the traction modules are provided with variable pitch rollers for allowing variable travel speed through the fuel channel. The cutting module is positioned in the middle of the device and it is equipped with three knife rolls for pressure tube cutting, using a system for cutting place video surveillance and pyrometers for monitoring cutting place temperature. The Cutting and Extraction Device (CED is fully automated, connected by wires to a Programmable Logic Controller (PLC and controlled from a Human Machine Interface (HMI. The design of the Cutting and Extraction Device (CED shall be achieved according to the particular features of the fuel channel components to be dismantled and to ensure radiation protection of workers.

  13. Experience on management of CANDU spent fuel in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H.-Y.; Choi, B.-I.; Yoon, J.-H.; Seo, U.-S. [Korea Hydro and Nuclear Power Co. Ltd., Nuclear Environment Technology Inst. (KHNP/NETEC), Yusung-Gu, Daejeon (Korea, Republic of)

    2002-07-01

    In Korea, national policy on the management of spent fuel from both PWR and CANDU reactors demands that all the spent fuel be kept within reactor site in until 2016 the time spent fuel interim storage facility might open. Based on the end of 2001, KHNP has 4 CANDU reactors in operation generating approximately 5,000 bundles of spent fuels per each unit annually. The generation, accumulation, and management of CANDU spent fuel by KHNP in Korea are reviewed. CANDU spent fuel storage technology including pool storage in fuel building, concrete silo storage, and on going project for consolidating storage adapting modular vault type MACSTOR concept are outlined. Especially current joint development of storage of CANDU spent fuel for improving land usage is addressed. The explanation of the new consolidated dry storage system includes description of the storage facility, its safety evaluations, and final implementation. Finally future movement on management of spent fuel in Korea is also briefly introduced. (author)

  14. The travesty of discarding used CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-15

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

  15. PARTICULARITIES REGARDING THE OPERATING PROCESS OF THE CUTTING AND EXTRACTION DEVICE IN THE CANDU HORIZONTAL FUEL CHANNELS PRESSURE TUBE DECOMMISSIONING PART II: CUTTING AND EXTRACTING PRESSURE TUBE PROCESS

    Directory of Open Access Journals (Sweden)

    Constantin POPESCU

    2016-05-01

    Full Text Available This paper presents some details of operation process for a Cutting and Extraction Device (CED in order to achieve the decommissioning of the horizontal fuel channels pressure tube in the CANDU 6 nuclear reactor. The most important characteristic of the Cutting and Extraction Device (CED is his capability of totally operator’s protection against the nuclear radiation during pressure tube decommissioning. The cutting and extracting pressure tube processes present few particularities due to special adopted technical solutions: a special module with three cutting rollers (system driven by an actuator, a guiding-extracting and connecting module (three fixing claws which are piloted by an actuator and block the device in the connecting position with extracting plugs. The Cutting and Extraction Device (CED is a train of modules equipped with special systems to be fully automated, connected with a Programmable Logic Controller (PLC and controlled by an operator panel type Human Machine Interface (HMI. All processes are monitored by video cameras. In case of error, the process is automatically stopped, the operator receiving an error message and the last sequence could be reinitialized or aborted due to safety reasons.

  16. Overview of activities on CANDU fuel in Argentina

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez, L.; Valesi, J., E-mail: lalvarez@cnea.gov.ar [National Commission on Atomic Energy, Fuel Engineering Department (Argentina)

    2013-07-01

    This paper gives an outline of activities on CANDU fuel in Argentina. It discusses the nuclear activities and electricity production in Argentina, evolution of the activities in fuel engineering, fuel fabrication, fuel performance at Embalse nuclear power plant and spent fuel storage options.

  17. Basic research and industrialization of CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-04-01

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

  18. Estimation of CANDU spent fuel disposal canister lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Kook, Dong Hak; Lee, Min Soo; Hwang, Yong Soo; Choi, Heui Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    Active nuclear energy utilization causes significant spent fuel accumulation problem. The cumulative amount of spent fuel is about 10,083 ton as of Dec. 2008, and is expected to increase up to 19,000 ton by 2020. Of those, CANDU spent fuels account for more than 60% of the total amounts. CANDU spent fuels had been stored in dry concrete silos since 1991 and during the past 15 years, 300 silos were constructed and {approx}3,200 ton of spent fuels are stored now. Another dry storage facility MACSTOR /KN-400 will store new-coming CANDU spent fuels from 2009. But, after intermediate storage ends, all CANDU spent fuels have to be disposed within multi-layer metallic canister which is composed of cast iron inside and copper outside. Canister lifetime estimation, therefore, is very important for the final disposal safety analysis. The most significant factor of lifetime is copper corrosion, and Y. S. Hwang developed a corrosion model in order to predict the general corrosion effect on copper canister lifetime during the final disposal period. This research applied his model to KURT1 where many disposal researches are being performed actively and the results shows safe margin of the copper canister for the very long-term disposal.

  19. Assessment of DUPIC fuel compatibility with CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

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

    Directory of Open Access Journals (Sweden)

    Constantin D. STANESCU,

    2015-05-01

    Full Text Available This paper presents a solution proposed by the authors in order to achieve of a cutting and extracting device operating panel for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The Cutting and Extraction Device (CED is fully automated, connected by wires to a Programmable Logic Controller (PLC and controlled from a Human Machine Interface (HMI. The Cutting and Extraction Device (CED performs the dismantling, cutting and extraction of the fuel channel components, moving with variable speed, temperature monitoring and video surveillance inside the pipe, unblock and extract the channel closure plug (from End Fitting - EF, unblock and extract the channel shield plug (from Lattice Tube - LT, block and cut the middle of the pressure tube, block and cut the end of the pressure tube, block and extract the half of pressure tube. All operations can be monitored and controlled from a operating panel. The PLC fully command the device in automatic or manually mode, to control the internal sensors, transducers, electrical motors, video surveillance and pyrometers for monitoring cutting place temperature. The device controller has direct access to the measured values with these sensors, interprets and processes them, preparing the next actionafter confirming the action in progress. The design of the Cutting and Extraction Device (CED shall be achieved according to the particular features of the fuel channel components to be dismantled and to ensure radiation protection of workers.

  1. A step towards closing the CANDU fuel cycle: an innovative scheme for reprocessing used CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Collins, F.; Lister, D. [Univ. of New Brunswick, UNB Nuclear, Dept. of Chemical Engineering, Fredericton, New Brunswick (Canada)

    2011-07-01

    Disposal versus reprocessing costs for used CANDU fuel was recently discussed by Rozon and Lister in a report produced for the Nuclear Waste Management Organization (NWMO). Their study discussed the economic incentives for reprocessing, not for the recovery of fissile uranium but for the recovery of plutonium ash. A $370/kg break-even price of uranium was calculated, and their model was found to be very sensitive to the reprocessing costs of the chosen technology. Findings were consistent with earlier studies done by Harvard University. Various reprocessing technologies (most based on solvent extraction) have been in use for many decades, but there appears to be no conceptual engineering study available in the open literature for a spent fuel reprocessing facility - one that includes process flows, operating costs and economic analysis. A deeper engineering study of the design and economics of re-processing technologies has since been undertaken by the nuclear group at the University of New Brunswick. An improved fluorination process was developed and modeled using ASPEN process simulation software. This study examines the impact of chosen technology on the spent fuel re-processing costs. (author)

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

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2015-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lau, J.H. [ed.

    1997-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

  6. Luncheon address: Early days of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Page, R.D. [Atomic Energy of Canada Limited (Canada)

    1997-07-01

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

  7. CURRENT STATUS OF INTEGRITY ASSESSMENT BY SIPPING SYSTEM OF SPENT FUEL BUNDLES IRRADIATED IN CANDU REACTOR

    Directory of Open Access Journals (Sweden)

    JONG-YOUL PARK

    2014-12-01

    Full Text Available In terms of safety and the efficient management of spent fuel storage, detecting failed fuel is one of the most important tasks in a CANada Deuterium Uranium (CANDU reactor operation. It has been successfully demonstrated that in a CANDU reactor, on-power failed fuel detection and location systems, along with alarm area gamma monitors, can detect and locate defective and suspect fuel bundles before discharging them from the reactor to the spent fuel storage bay. In the reception bay, however, only visual inspection has been used to identify suspect bundles. Gaseous fission product and delayed neutron monitoring systems cannot precisely distinguish failed fuel elements from each fuel bundle. This study reports the use of a sipping system in a CANDU reactor for the integrity assessment of spent fuel bundles. The integrity assessment of spent fuel bundles using this sipping system has shown promise as a nondestructive test for detecting a defective fuel bundle in a CANDU reactor.

  8. The small (or large) modular CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Meneley, D.; Harvel, G. [Univ. of Ontario Inst. of Tech., Oshawa, Ontario (Canada)

    2013-07-01

    This presentation outlines the design for small (or large) modular CANDU. The origins of this work go back many years to a comment by John Foster, then President of AECL CANDU. Foster noted that the CANDU reactor, with its many small fuel channels, was like a wood campfire. To make a bigger fire, just throw on some more logs (channels). If you want a smaller fire, just use fewer logs. The design process is greatly simplified.

  9. Evaluation of safety margins during dry storage of CANDU fuel in MACSTOR/KN-400 module

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, R.; Shill, R. [Atomic Energy Of Canada Limited, Montreal, Quebec (Canada); Lee, K.-H.; Chung, S.-H.; Yoon, J.-H.; Choi, B.-I.; Lee, H.-Y.; Song, M.-J. [KHNP, Nuclear Environment Technology Inst., Taejon (Korea, Republic of)

    2005-03-15

    This paper covers an evaluation of the available safety margin against fuel bundle degradation during dry storage of CANDU spent fuel bundles in a MACSTOR/KN-400 module, considering normal, off-normal and postulated accidental conditions. (author)

  10. Procurement and supply of CANDU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Bazeley, E.G. [E.G. Bazeley and Associates, Whitby, Ontario (Canada)

    2002-11-01

    In 1955 a decision was made to proceed with construction of a Nuclear Power Demonstration Station (NPD) near Rolfton, Ontario. This project, headed by Atomic Energy of Canada with major involvement of private industry, was the genesis for the development of nuclear electric generation in Canada. This paper reviews one aspect of the Canadian program: the evolution of fuel procurement and supply, which in itself has been a remarkable Canadian achievement. (author)

  11. ENHANCING ADVANCED CANDU PROLIFERATION RESISTANCE FUEL WITH MINOR ACTINIDES

    Energy Technology Data Exchange (ETDEWEB)

    Gray S. Chang

    2010-05-01

    The advanced nuclear system will significantly advance the science and technology of nuclear energy systems and to enhance the spent fuel proliferation resistance. Minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs can play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In this work, an Advanced CANDU Reactor (ACR) fuel unit lattice cell model with 43 UO2 fuel rods will be used to investigate the effectiveness of a Minor Actinide Reduction Approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance. The main MARA objective is to increase the 238Pu / Pu isotope ratio by using the transuranic nuclides (237Np and 241Am) in the high burnup fuel and thereby increase the proliferation resistance even for a very low fuel burnup. As a result, MARA is a very effective approach to enhance the proliferation resistance for the on power refueling ACR system nuclear fuel. The MA transmutation characteristics at different MA loadings were compared and their impact on neutronics criticality assessed. The concept of MARA, significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate term of nuclear energy reconnaissance.

  12. The mode of operation of CANDU power reactor in thorium self-sufficient fuel cycle

    Directory of Open Access Journals (Sweden)

    Bergelson Boris R.

    2008-01-01

    Full Text Available This paper presents the results of calculations for CANDU reactor operation in the thorium fuel cycle. The calculations were performed to estimate feasibility of operation of a heavy-water thermal neutron power reactor in the self-sufficient thorium cycle. The parameters of the active core and the scheme of fuel reloading were considered to be the same as for the standard operation in the uranium cycle. Two modes of operation are discussed in the paper: the mode of preliminary accumulation of 233U and the mode of operation in the self-sufficient cycle. For calculations for the mode of accumulation of 233U, it was assumed that plutonium was used as the additional fissile material to provide neutrons for 233U production. Plutonium was placed in fuel channels, while 232Th was located in target channels. The maximum content of 233U in the target channels was about 13 kg/t of ThO2. This was achieved by six year irradiation. The start of reactor operation in the self-sufficient mode requires content of 233U not less than 12 kg/t. For the mode of operation in the self-sufficient cycle, it was assumed that all the channels were loaded with the identical fuel assemblies containing ThO2 and a certain amount of 233U. It was shown that the non-uniform distribution of 233U in a fuel assembly is preferable.

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

    Indian Academy of Sciences (India)

    B R Bergelson; A S Gerasimov; G V Tikhomirov

    2007-02-01

    This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the same as for standard operation in uranium cycle. Two modes of operations are discussed in the paper: mode of preliminary accumulation of 233U and mode of operation in self-sufficient cycle. For the mode of accumulation of 233U it was assumed for calculations that plutonium can be used as additional fissile material to provide neutrons for 233U production. Plutonium was placed in fuel channels, while 232Th was located in target channels. Maximum content of 233U in target channels was estimated to be ∼ 13 kg/t of ThO2. This was achieved by irradiation for six years. The start of the reactor operation in the self-sufficient mode requires 233U content to be not less than 12 kg/t. For the mode of operation in self-sufficient cycle, it was assumed that all channels were loaded with identical fuel assemblies containing ThO2 and certain amount of 233U. It is shown that nonuniform distribution of 233U in fuel assembly is preferable.

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

    Energy Technology Data Exchange (ETDEWEB)

    Marino, A.C. [Comison Nacional de Energia Atomica (Argentina)

    1997-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-09-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  20. Automated refueling simulations of a CANDU for the exploitation of thorium fuels

    Science.gov (United States)

    Holmes, Bradford

    CANDU nuclear reactors are in a unique circumstance where they are able to utilize and exploit a number of different fuel options to provide power as a utility. Thorium, a fertile isotope found naturally, is one option that should be explored. Thorium is more abundant than uranium, which is the typical fuel in the reactor and the availability of thorium makes nuclear energy desirable to more countries. This document contains the culmination of a project that explores, tests, and analyzes the feasibility of using thorium in a CANDU reactor. The project first develops a set of twodimensional lattice and three dimensional control rod simulations using the DRAGON Version 4 nuclear physics codes. This step is repeated for many concentrations of thorium. The data generated in these steps is then used to determine a functional enrichment of thorium. This is done via a procedural elimination and optimization of certain key parameters including but not limited to average exit burnup and reactivity evolution. For the purposes of this project, an enrichment of 1 % thorium was found viable. Full core calculations were done using the DONJON 4 code. CANFUEL, a program which simulates the refueling operations of a CANDU reactor for this fuel type was developed and ran for a simulation period of one hundred days. The program and the fuel selection met all selected requirements for the entirety of the simulation period. CANFUEL requires optimization for fuel selection before it can be used extensively. The fuel selection was further scrutinized when a reactivity insertion event was simulated. The adjuster rod 11 withdrawal from the core was analyzed and compared to classical CANDU results in order to ensure no significant deviations or unwanted evolutions were encountered. For this case, the simulation results were deemed acceptable with no significant deviations from the classical CANDU case.

  1. R and D activities on CANDU-type fuel in Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Suripto, A.; Badruzzaman, M.; Latief, A. [Nuclear Fuel Element Centre, National Atomic Energy Agency of Indonesia (BATAN), Puspiptek, Serpong (Indonesia)

    1997-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-08-15

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

  3. A controllability study of TRUMOX fuel for load following operations in a CANDU-900 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Trudell, D.A., E-mail: trudelda@mcmaster.ca [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    Using a core model of a generic CANDU-900 reactor in RFSP-IST, load following simulations have been performed to assess the controllability of the reactor due to Xenon transients. Week long load following simulations have been performed with daily power cycles 12 hours in duration. Simulations have shown that Natural Uranium fuel can be safely cycled between 100 and 90% Full Power without adjuster rod movement while TRUMOX fuel can be safely cycled between 100 and 85% Full Power. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. CANDU fuel attribution through the analysis of delayed neutron temporal behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Sellers, M.T.; Corcoran, E.C.; Kelly, D.G., E-mail: David.Kelly@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)

    2012-07-01

    Delayed Neutron Counting (DNC) is an established technique in the Canadian nuclear industry as it is used for the detection of defective fuel in several CANDU reactors and the assay of uranium in geological samples. This paper describes the possible expansion of DNC to the discipline of nuclear forensics analysis. The temporal behaviour of experimentally measured delayed neutron spectra were used to determine the relative contributions of {sup 233}U and {sup 235}U to the overall fissile content present in mixtures with average absolute errors of ±4 %. The characterization of fissile content in current and proposed CANDU fuels (natural UO{sub 2}, thoria and mixed oxide (MOX) based) by DNC analysis is evaluated through Monte Carlo simulations. (author)

  6. Measurement of gap and grain-boundary inventories of {sup 129}I in used CANDU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Stroes-Gascoyne, S.; Moir, D.L.; Kolar, M.; Porth, R.J.; McConnell, J.L.; Kerr, A.H. [AECL Research, Pinawa, Manitoba (Canada). Whiteshell Labs.

    1995-12-31

    Combined gap and grain-boundary inventories of {sup 129}I in 14 used CANDU fuel elements were measured by crushing and simultaneously leaching fuel segments for 4 h in a solution containing KI carrier. From analogy with previous work a near one-to-one correlation was anticipated between the amount of stable Xe and the amount of {sup 128}I in the combined gap and grain-boundary regions of the fuel. However, the results showed that such a correlation was only apparent for low linear power rating (LLPR) fuels with an average linear power rating of < 42 kW/m. For high linear power rating (HLPR) fuels (> 44 kW/m), the {sup 129}I values were considerably smaller than expected. The combined gap and grain-boundary inventories of {sup 129}I in the 14 fuels tested varied from 1.8 to 11.0%, with an average value of 3.6 {+-} 2.4% which suggests that the average value of 8.1 {+-} 1% used in safety assessment calculations overestimates the instant release fraction for {sup 129}I. Segments of used CANDU fuels were leached for 92 d (samples taken at 5, 28 and 92 d) to determine the kinetics of {sup 129}I release. Results could be fitted tentatively to half-order reaction kinetics, implying that {sup 129}I release is a diffusion-controlled process for LLPR fuels, and also for HLPR fuels, once the gap inventory has been leached. However, more data are needed over longer leaching periods to gain more understanding of the processes that control grain-boundary release of {sup 129}I from used CANDU fuel.

  7. FAST: A Fuel And Sheath Modeling Tool for CANDU Reactor Fuel

    Science.gov (United States)

    Prudil, Andrew Albert

    before for any Canadian fuel performance code). This thesis documents the theory employed by the model, its implementation, and the results of a proof of concept validation. The validation compared model predictions against both experimental data and results obtained from the ELESTRES and ELOCA fuel performance codes. Overall, the results show excellent model performance except in cases of a strong axial dependence. An analysis of the sensitivity of the model to the uncertainty in input parameters and the material properties is also presented. Finally, this thesis includes a discussion of the limitations, applications, and potential for future development of code. Key words: nuclear fuel, CANDU fuel, fuel modeling, multiphysics modeling, Comsol

  8. Development of a System Dynamics Model for Evaluating the Economics of an Advanced CANDU Fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-15

    Since the early 1990's, the Korea Atomic Energy Research Institute (KAERI) and the Atomic Energy of Canada Limited (AECL) have cooperated to develop, verify, and demonstrate the advanced CANDU fuel, so called CANFLEX-NU (Natural Uranium). The CANFLEX-NU fuel bundle consists of 43 fuel elements and has the buttons on the outer surface of the fuel elements for improving the CHF (Critical-Heat-Flux) characteristics. Because of this features of CANFLEXNU fuel, it offers higher operating and safety margins than current 37-element fuel. Recently, the interest for a CANFLEX-NU has been increased because of the power de-rating due to aging of CANDU reactors. Wolsong Unit 1 CANDU reactor has been operated over 25 years and the operating power at the present time is less than 90% of a full power because of a reduction of the margin of ROP trip set point. The most appropriate way to overcome such a power de-rating due to a crept pressure tube is the introduction of a CANFLEX-NU fuel into a CANDU reactor. Now, a CANFLEX-NU fuel is ready to be commercialized in a CANDU-6 reactor because the design and demonstration irradiation have been completed in both Korea and Canada. Economic evaluation for commercializing a CANFLEX-NU fuel in Wolsong Units was carried out by calculating the unit prime cost of electricity production. Throughout the economic evaluation, it was found that the introduction of CANFLEX-NU fuel into Wolsong Units would have much economic benefits due to a better operating performance. However, the amount of economic profit due to introducing CANFLEX-NU fuel depends on several parameters such as the required time to get license from regulatory institute before commercializing, licensing cost, failure probability of commercializing etc. Therefore, it is necessary to determine the optimum condition to get the highest economic profit. In this paper, an economic evaluation was carried out based on the starting year of the licensing study with considering the

  9. Waste management issues and their potential impact on technical specifications of CANDU fuel materials

    Energy Technology Data Exchange (ETDEWEB)

    Tait, J.C.; Johnson, L.H. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada)

    1997-07-01

    The technical specifications for the composition of nuclear fuels and materials used in Canada's CANDU reactors have been developed by AECL and materials manufacturers, taking into account considerations specific to their manufacture and the effect of minor impurities on fuel behaviour in reactor. Nitrogen and chlorine are examples of UO{sub 2} impurities, however, where there is no technical specification limit. These impurities are present in the source materials or introduced in the fabrication process and are neutron activated to {sup 14}C and {sup 36}C1, which after {sup 129}I , are the two most significant contributors to dose in safety assessments for the disposal of used fuel. For certain impurities, environmental factors, particularly the safety of the disposal of used fuels, should be taken into consideration when deriving 'allowable' impurity limits for nuclear fuel materials. (author)

  10. Design requirements of a consolidating dry storage module for CANDU spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Ho; Yoon, Jeong Hyoun; Yang, Ke Hyung; Choi, Byung Il; Lee, Heung Young [KHNP/NETEC, Taejon (Korea, Republic of); Cho, Gyu Seong [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2003-10-01

    This paper presents a technical description of design requirement document covers the requirements of the MACSTOR/KN-400 module, which is under development to densely accommodate CANDU spent fuels with more efficient way. The design requirement is for the module that will be constructed within a dry storage site after successfully licensed by the regulatory body. This temporary outdoor spent fuel dry storage facility provides for safe storage of spent nuclear fuel after it has been removed from the plant's storage pool after being allowed to decay for a period of at least 6 years. The MACSTOR/KN-400 module is being designed to the envelope of site environmental conditions encountered at the Wolsong station. The design requirements of MACSTOR/KN-400 module meets the requirements of the appropriate Codes and Standards for dry storage of spent fuel from nuclear power reactors such as lOCFR72, and Korea Atomic Energy Act and relevant technical standard.

  11. Evaluation of safety margins during dry storage of CANDU fuel in MACSTOR/KN-400 module

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, R.; Shill, R. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada); Lee, K.-H.; Chung, S.-H.; Yoon, J.-H.; Choi, B.-I.; Lee, H.-Y.; Song, M.-J. [KHNP, Nuclear Environment Technology Inst., Taejon (Korea, Republic of)

    2004-07-01

    This paper covers an evaluation of the available safety margin against fuel bundle degradation during dry storage of CANDU spent fuel bundles in a MACSTOR/KN-400 module, considering normal, off-normal and postulated accidental conditions. Korea Hydro and Nuclear Power (KHNP), in collaboration with Atomic Energy of Canada Limited (AECL), are developing a new module for the dry storage of spent fuel from the four CANDU 6 nuclear reactors at the Wolsong site in South Korea. The module provides the benefit of occupying significantly less area than the concrete canisters presently used. The modules are designed for a minimum service life of 50 years. During that period, the spent fuel bundles shall be safely stored. This imposes that failure of a fuel bundle element or unacceptable degradation of an existing defect (from reactor operation) does not occur during the dry storage period. The fuel bundles are stored in an air-filled fuel basket that releases 365 Watts on average and a maximum of 390 Watts when rare fuel loading conditions are postulated. In addition, specific accidental air flow cooling conditions are postulated that consist of 100% blockage of all air inlets on one side of the module. These conditions can generate a peak daily fuel temperature of up to 155{sup o}C during a reference hot summer day during the first year of operation. The fuel temperature decreases over the years and also fluctuates due to daily and seasonal temperature variations. At this temperature, fuel elements with intact Zircaloy sheathing will not experience damage. However, for the few fuel bundle elements that are non-leaktight (less than 1 per 37,000), some re-oxidation of UO{sub 2} into higher oxides such as U{sub 3}O{sub 7} / U{sub 4}O{sub 9} and U{sub 3}O{sub 8} will occur. This latter form of Uranium oxide is undesirable due to its lower density that results in a volumetric increase of the pellet that can overstress the fuel element sheathing. The level of fuel pellet

  12. FGR Evaluation Code for CANDU Fuel under an Accidents: REDOU

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jung, Jong Yeob

    2006-12-15

    When an end fitting failure is occurred, the sheath is broken and the fission gases are released promptly from the gap between the pellet and sheath and also released continuously from the inside of the pellet due to the oxidation of the pellet. Thus, the accurate calculation of the fission gas release from the gap and from the inside of the pellet is an essential to the accident analysis of the fuel behavior and to prepare the safety strategy. In this report, among the performance analysis or the transient behavior prediction computer codes, the REDOU code which is the fission gas calculating code for the postulated accident scenario such as an end fitting failure is introduced. Then, the user manual for REDOU code is provided so that it can be the guidance to the potential users of the code and save the time and economic loss by reducing the trial and error.

  13. Thermal-hydraulic interfacing code modules for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  14. Analytical assessment for stress corrosion fatigue of CANDU fuel elements under load following conditions

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore; Ionescu, Drags; Pauna, Eduard [Institute for Nuclear Research, Pitesti (Romania). Nuclear Fuel Engineering Lab.

    2012-03-15

    When nuclear power reactors are operated in a load following (LF) mode, the nuclear fuel may be subjected to step changes in power on weekly, daily, or even hourly basis, depending on the grid's needs. Two load following tests performed in TRIGA Research Reactor of Institute for Nuclear Research (INR) Pitesti were simulated with finite elements computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the sheath arising from expansion and contraction of the pellets in the corrosive environment. The 3D finite element analyses show that the cyclic strains give highly multiaxial stresses in the sheath at ridge region. This paper summarizes the results of the analytical assessment for SCF and their relation to CANDU fuel performance in LF tests conditions. (orig.)

  15. Optimization of the self-sufficient thorium fuel cycle for CANDU power reactors

    Directory of Open Access Journals (Sweden)

    Bergelson Boris R.

    2008-01-01

    Full Text Available The results of optimization calculations for CANDU reactors operating in the thorium cycle are presented in this paper. Calculations were performed to validate the feasibility of operating a heavy-water thermal neutron power reactor in a self-sufficient thorium cycle. Two modes of operation were considered in the paper: the mode of preliminary accumulation of 233U in the reactor itself and the mode of operation in a self-sufficient cycle. For the mode of accumulation of 233U, it was assumed that enriched uranium or plutonium was used as additional fissile material to provide neutrons for 233U production. In the self-sufficient mode of operation, the mass and isotopic composition of heavy nuclei unloaded from the reactor should provide (after the removal of fission products the value of the multiplication factor of the cell in the following cycle K>1. Additionally, the task was to determine the geometry and composition of the cell for an acceptable burn up of 233U. The results obtained demonstrate that the realization of a self-sufficient thorium mode for a CANDU reactor is possible without using new technologies. The main features of the reactor ensuring a self-sufficient mode of operation are a good neutron balance and moving of fuel through the active core.

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

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, F.; Hadaller, G.I.; Fortman, R.A., E-mail: fabbasian@sternlab.com [Stern Laboratories Inc., Hamilton, Ontario (Canada)

    2013-07-01

    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)

  17. Leaching of used CANDU fuel: Results from a 19-year leach test under oxidizing conditions

    Energy Technology Data Exchange (ETDEWEB)

    Stroes-Gascoyne, S.; Johnson, L.H.; Tait, J.C.; McConnell, J.L.; Porth, R.J. [AECL, Pinawa, Manitoba (Canada). Whiteshell Labs.

    1997-12-31

    A fuel leaching experiment has been in progress since 1977 to study the dissolution behavior of used CANDU fuel in aerated aqueous solution. The experiment involves exposure of 50-mm clad segments of an outer element of a Pickering fuel bundle (burnup 610 GJ/kg U; linear and peak power ratings 53 and 58 kW/m, respectively), to deionized distilled water (DDH{sub 2}O, {approximately}2 mg/L carbonate) and tapwater ({approximately}50 mg/L carbonate). In 1992, it was observed that the fuel in at least one of the leaching solutions showed some signs of deterioration and, therefore, in 1993, parts of the fuel samples were sacrificed for a detailed analysis of the physical state of the fuel, using SEM and optical microscopy. Leaching results to date show that even after >6900 days only 5 to 7.7% of the total calculated inventory of {sup 137}Cs has leached out preferentially and that leach rates suggest a development towards congruent dissolution. Total amounts of {sup 137}Cs and {sup 90}Sr leached are slightly larger in tapwater than in DDH{sub 2}O. SEM examinations of leached fuel surface fragments indicate that the fuel surface exposed to DDH{sub 2}O is covered in a needle-like precipitate. The fuel surface exposed to tapwater shows evidence of leaching but no precipitate, likely because uranium is kept in solution by carbonate. Detailed optical and SEM microscopy examinations on fuel cross sections suggest that grain-boundary dissolution in DDH{sub 2}O is not prevalent, and in tapwater appears to be limited to the outer {approximately}0.5 mm (pellet/cladding) region of the fuel. Grain boundary attack seems to be limited to microcracks at or near the surface of the fuel. It thus appears that grain-boundary attack occurs only near the fuel pellet surface and is prevalent only in the presence of carbonate in solution.

  18. Structural design concept and static analysis of CANDU spent fuel compact dry storage system

    Energy Technology Data Exchange (ETDEWEB)

    Choi, K. S.; Yang, K. H.; Paek, C. R.; Jung, J. S.; Lee, H. Y. [Korea Hydro and Nuclear Power Company, Taejon (Korea, Republic of)

    2003-07-01

    In this study, an structural design concept on CANDU spent fuel compact dry storage system MACSTOR/KN-400 module has been established with a view to optimally design the structural members of the system. Design loads, loading combination and structural safety criteria of the module were reviewed assuming W olsung Site. The static analysis of the module showed that compressive stress concentration due to dead load and live load occurred around the center of roof slab. Maximum stress resulted from dead load is about twice as much as the stress from live load, and structural behavior of module caused by wind load was not significant. The static analysis results will have influence on the reinforcement bar design of structural members with other structural analyses.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  20. Localization of CANDU technology

    Energy Technology Data Exchange (ETDEWEB)

    Alizadeh, Ala

    2010-09-15

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

  1. Seismic Soil-Structure Interaction Analysis of a Consolidated Dry Storage Module for CANDU Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Gon; Yoon, Jeong Hyoun; Cho, Chun Hyung; Lee, Heung Young [Korea Hydro and Nuclear Power Co., Ltd., Taejon (Korea, Republic of); Choi, Kyu Sup; Jeong, In Su; Kim, Jong Soo [KONES Co., Seoul (Korea, Republic of)

    2005-07-01

    The MACSTOR/KN-400 module has been developed as an effective alternative to the existing stand alone concrete canister for dry storage of CANDU spent fuel. The structure is a concrete monolith of 21.67 m long and 12.66 m wide and has a height equal to 7.518 m including the bottom slab. Inside of the concrete module are built 40 storage cylinders accommodating ten 60- bundle dry storage baskets, which are suspended from the top slab and eventually constrained at 10 cm above the bottom slab with horizontal seismic restraints. The main cooling process of the MACSTOR/KN-400 module is by air convection through air inlets and outlets. The civil design parameters, with respect to meteorological and seismic loads applied to the module are identical to those specified for the Wolsung CANDU 3 and 4 plants except for local geologic characteristics. As per USNRC SRP Section 3.7.2 and current US practices, Soil-Structure Interaction (SSI) effect shall be considered for all structures not supported by a rock or rock-like soil foundation materials. An SSI is a very complicated phenomenon of the structure coupled with the soil medium that is usually semi-infinite in extent and highly nonlinear in its behavior. And the effect of the SSI is noticeable especially for stiff and massive structures resting on relatively soft ground. Thus the SSI effect has to be considered in the seismic design of MACSTOR/KN-400 module resting on soil medium. The scope of the this paper is to carry out a seismic SSI analysis of the MACSTOR/KN-400 module, in order to show how much the SSI gives an effect on the structural responses by comparing with the fixed-base analysis.

  2. Current Status and Characterization of CANDU Spent Fuel for Geological Disposal System Design

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Dong Keun; Lee, Seung Woo; Cha, Jeong Hun; Choi, Jong Won; Choi, Heui Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Yang [SK Engineering and Construction, Seoul (Korea, Republic of)

    2008-06-15

    Inventories to be disposed of, reference turn up, and source terms for CANDU spent fuel were evaluated for geological disposal system design. The historical and projected inventory by 2040 is expected to be 14,600 MtU under the condition of 30-year lifetime for unit 1 and 40-year lifetime for other units in Wolsong site. As a result of statistical analysis for discharge burnup of the spent fuels generated by 2007, average and stand deviation revealed 6,987 MWD/MtU and 1,167, respectively. From this result, the reference burnup was determined as 8,100 MWD/MtU which covers 84% of spent fuels in total. Source terms such as nuclide concentration for a long-term safety analysis, decay heat, thermo-mechanical analysis, and radiation intensity and spectrum was characterized by using ORIGEN-ARP containing conservativeness in the aspect of decay heat up to several thousand years. The results from this study will be useful for the design of storage and disposal facilities.

  3. Drop Test of the Candu Spent Fuel Storage Basket in MACSTOR/KN-400

    Energy Technology Data Exchange (ETDEWEB)

    Choi, W.S.; Jeon, J.Y.; Seo, K.S. [KAERI, 1045 Daedeokdaero, Yuseong, Daejeon, 305-353 (Korea, Republic of); Park, J.E.; Yoo, G.S.; Park, W.G. [Korea Hydro Nuclear Power - KHNP (Korea, Republic of)

    2009-06-15

    The MACSTOR/KN-400 of Wolsung power plant in Korea is a dry interim storage facilities. There are 400 long slender cylinders in MACSTOR/KN-400. In one cylinder, ten baskets where Candu spent fuels are loaded are stacked and stored. For this MACSTOR/KN-400 facilities, analyses and tests for the hypothetical accident conditions that might happen during moving and storing baskets into a cylinder were performed. The hypothetical accident conditions to be considered are two cases. One is the case of basket dropping onto the bottom plate of a cylinder. The other is the case of basket dropping onto the other basket top plate stored in the cylinder. For the drop analyses, the case of hanging cylinder and the case of cylinder on the unyielding target surface were considered. Based on the dropping analysis, testing condition was determined as the latter case that is for the cylinder on the target surface. In a basket, 60 dummy fuel bundles are loaded which have the same weight of real spent fuel bundles. On the external surface of the basket, 8 strain gauges and 4 accelerometers were attached for the data acquisition. In order to measure the velocity when a basket impacts, three different devices were utilized. And the impact velocity results were compared and cross-checked. After the dropping tests, helium leak tests were conducted to evaluate the leakage rate. (authors)

  4. A comprehensive model for in-plane and out-of-plane vibration of CANDU fuel endplate rings

    Energy Technology Data Exchange (ETDEWEB)

    Yu, S.D., E-mail: syu@ryerson.ca; Fadaee, M.

    2016-08-01

    Highlights: • Proposed an effective method for modelling bending and torsional vibration of CANDU fuel endplate rings. • Applied successfully the thick plate theory to curved structural members by accounting for the transverse shear effect. • The proposed method is computationally more efficient compared to the 3D finite element. - Abstract: In this paper, a comprehensive vibration model is developed for analysing in-plane and out-of-plane vibration of CANDU fuel endplate rings by taking into consideration the effects of in-plane extension in the circumferential and radial directions, shear, and rotatory inertia. The model is based on Reddy’s thick plate theory and the nine-node isoparametric Lagrangian plate finite elements. Natural frequencies of various modes of vibration of circular rings obtained using the proposed method are compared with 3D finite element results, experimental data and results available in the literature. Excellent agreement was achieved.

  5. Simulation of transient heat transfer in MACSTOR/KN-400 module storing irradiated CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sabourin, G. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada); Lee, K.-H.; Yoon, J.-H.; Choi, B.-I.; Lee, H.-Y.; Song, M.-J. [KHNP, Nuclear Environment Technology Inst., Taejon (Korea, Republic of)

    2004-07-01

    Korea Hydro and Nuclear Power (KHNP), in collaboration with Atomic Energy of Canada Limited (AECL), are developing a new module for the dry storage of spent fuel from the four CANDU 6 nuclear reactors at the Wolsong site in South Korea, the MACSTOR/KN-400. The simulation of transient conditions for AECL's spent fuel dry storage systems, presented in this paper, has not been performed before and is considered a major achievement of the present work. In a fist step, CATHENA was compared to MACSTOR-200 temperature measurements and the accuracy of the results were very good. In a second step, CATHENA was applied to the MACSTOR/KN-400. Four cases were performed for the MACSTOR/KN-400: Off-normal cases in summer and winter and reduced air flow cases in summer and winter. The maximum local concrete temperatures were predicted to be 63{sup o}C for the off-normal case and 65{sup o}C in the reduced air flow case. The maximum temperature gradients in the concrete are predicted to be 28{sup o}C for the off-normal case and 30{sup o}C in the reduced air flow case, incorporating a 3{sup o}C uncertainty. This paper shows that the maximum temperature for the module is expected to meet the temperature limitations of appropriate standards. (author)

  6. Heat transfer analysis of the MACSTOR/KN-400 storage module for CANDU spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Youn, J. H.; Choi, B. I.; Lee, H. Y. [Nuclear Environment Technology Institute, Taejon (Korea, Republic of)

    2003-10-01

    It was verified through heat transfer analysis that a consolidated dry storage system for CANDU spent fuel, MACSTOR/KN-400 was safe in thermal aspect. In order to validate the computer code of CATHENA which was employed to perform the analysis, the comparison between actual measurement data of MACSTOR-200 at Getilly-2 NPP in Canada and computed values from the code has been carried out. The comparison represented that the computed values acceptably agreed to the measurement data and thus the computer code was verified for its application to MACSTOR/KN-400. The identical K-values(parameter to describe head loss inside the module) and convective heat transfer coefficient of the module obtained by the validation was applied to the heat transfer analysis modelling of MACSTOR/KN-400. The result from the analysis showed that under 40 .deg. C of ambient temperature, maximum average and local temperatures of the concrete module were represented by 53 .deg. C and 69 .deg. C, respectively, which fulfilled well the allowable temperature limit of the concrete structure given by ACI349(American Concrete Institute)

  7. Proceedings of the fourth international conference on CANDU maintenance

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

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

  8. Release of [sup 14]C from the gap and grain-boundary regions of used CANDU fuels to aqueous solutions

    Energy Technology Data Exchange (ETDEWEB)

    Stroes-Gascoyne, S.; Tait, J.C.; Porth, R.J.; McConnell, J.L.; Lincoln, W.J. (Whiteshell Lab., Pinawa, Manitoba (Canada). AECL Research)

    1994-01-01

    This study was undertaken as part of the Canadian Nuclear Fuel Waste Management Program (CNFWMP), to measure [sup 14]C inventories of used CANDU fuel. Other objectives were to measure the fraction of the total [sup 14]C inventory that would be instantly released to solution from used CANDU fuels upon sheath failure and to determine if the assumptions made in safety assessment calculations of used fuel waste disposal regarding instant release of [sup 14]C were correct. Results showed that the measured [sup 14]C inventories were a factor of 11.5 [+-] 3.9 lower than the estimated [sup 14]C inventory values used in safety assessment calculations. Measured instant release values for [sup 14]C ranged from 0.06 to 5.04% (of total [sup 14]C inventories) with an average of 2.7 [+-] 1.6%, indicating that instant release fractions for [sup 14]C used in safety assessment calculations (1.2--25%) were overestimated.

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Choi, Hang Bok

    2005-03-01

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

  10. FAST: A combined NOC and transient fuel model for CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Prudil, A.; Lewis, B.J.; Chan, P.K., E-mail: Paul.Chan@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada); Baschuk, J.J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    The Fuel And Sheath modelling Tool (FAST) is a fuel performance code that is being developed for both normal and transient operating conditions. FAST includes models for heat generation and transport, thermal-expansion, elastic strain, densification, fission product swelling, pellet relocation, contact, grain growth, fission gas release, gas and coolant pressure and sheath creep. These models have been implemented using the Comsol finite-element platform. The equations are solved on a two-dimensional (radial-axial) geometry of a fuel pellet and sheath. FAST has undergone a proof of concept validation against experimental data and comparison to the ELESTRES and ELOCA fuel performance codes. The results show excellent agreement with experimental measurements and the above stated IST- codes. (author)

  11. Feasibility study in aspect of thermal integrity on the dry storage expansion options for CANDU spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Yoon, J. H.; Choi, B. I.; Lee, H. Y.; Song, M. J. [Nuclear Environment Technology Institute, Taejon (Korea, Republic of); Cho, K. S. [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2002-10-01

    In order to expand the capability of the CANDU spent fuel dry storage facilities of the at Wolsong, the alternative concepts based on MACSTOR are suggested to replace with existing concrete silo of Wolsong. For this, the feasibility of its design changes from original MACSTOR is examined in term of heat transfer and thermal hydraulic. In this study, the configuration of the module was conceptually changed from its original 2 rows to 3 and 4 rows for review. Under normal operation, the results of heat transfer and thermal hydraulic shows that storage module can feasibly accomodate four rows of storage cylinders within allowable range in terms of maximum allowable temperature of the fuel basket.

  12. A Deformation Analysis Code of CANDU Fuel under the Postulated Accident: ELOCA

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jung, Jong Yeob

    2006-11-15

    Deformations of the fuel element or fuel channel might be the main cause of the fuel failure. Therefore, the accurate prediction of the deformation and the analysis capabilities are closely related to the increase of the safety margin of the reactor. In this report, among the performance analysis or the transient behavior prediction computer codes, the analysis codes for deformation such as the ELOCA, HOTSPOT, CONTACT-1, and PTDFORM are briefly introduced and each code's objectives, applicability, and relations are explained. Especially, the user manual for ELOCA code which is the analysis code for the fuel deformation and the release of fission product during the transient period after the postulated accidents is provided so that it can be the guidance to the potential users of the code and save the time and economic loss by reducing the trial and err000.

  13. Reactor Physics Analysis Models for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok

    2007-10-15

    Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.

  14. Development of CANDU advanced fuel fabrication technology - A development of amorphous alloys for the solder of nuclear reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jai Young; Lee, Ki Young; Kim, Yoon Kee; Jung, Jae Han; Yu, Ji Sang; Kim, Hae Yeol; Han, Young Su [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1996-07-01

    In the case of advanced CANDU fuel being useful in future, the fabrication processes for soundness insurance of a improved nuclear fuel bundle must be developed at the same time because it have three times combustibility as existing fuel. In particular, as the improved nuclear fuel bundle in which a coated layer thickness is thinner than existing that, firmity of a joint part is very important. Therefore, we need to develop a joint technique using new solder which can settle a potential problem in current joining method. As the Zr-Be alloy system and the Ti-Be system are composed with the elements having high neutron permeability, they are suitable for joint of nuclear fuel pack. The various compositions Zr-Be and Ti-Be binary metallic glass alloys were applicable to the joining the nuclear fuel bundles. The thickness of joint layer using the Zr{sub 1-x} Be{sub x} amorphous ribbon as a solder is thinner than that using physical vapor deposited Be. Among the Zr{sub 1-x} Be{sub x} amorphous binary alloys, Zr{sub 0.7} Be{sub 0.3} binary alloy is the most appropriated for joint of nuclear fuel bundle because its joint layer is smooth and thin due to low degree of Be diffusion. The microstructures of brazed layer using Ti{sub 1-y} Be{sub y} alloy, however, a solid-solution layer composed with Zr and Ti is formed toward the Zr cladding sheath and many of Zr is detected in the joint lever. 20 refs., 8 tabs., 23 figs. (author)

  15. Pre-licensing of the Advanced CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ion, R.; Popov, N.; Snell, V.; Xu, C. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); West, J. [Candesco Co., Toronto, Ontario (Canada)

    2006-09-15

    Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving 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 electrical output of 1165 MWe. The ACR-1000 design has evolved from AECL's in-depth knowledge of CANDU systems, components, and materials, as well as the experience and feedback received from owners and 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. It also features major improvements in economics, inherent safety characteristics, and performance, while retaining the proven benefits of the CANDU family of nuclear power plants. The CANDU system is ideally suited to this evolutionary approach since the modular fuel channel reactor design can be modified, through a series of incremental changes in the reactor core design, to increase the power output and improve the overall safety, economics, and performance. The safety enhancements made in ACR-1000 encompass improved safety margins, performance and reliability of safety related systems. In particular, the use of the CANFLEX-ACR fuel bundle, with lower linear rating and higher critical heat flux, provides increased operating and safety margins. Safety features draw from those of the existing CANDU plants (e.g., the two

  16. Scenarios for the transmutation of actinides in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hyland, Bronwyn, E-mail: hylandb@aecl.ca [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada); Gihm, Brian, E-mail: gihmb@aecl.ca [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (Canada)

    2011-12-15

    With world stockpiles of used nuclear fuel increasing, the need to address the long-term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes to reduce the decay heat period or consumed as fuel in a CANDU(R) reactor. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle simplifies the fabrication and handling of active fuels. Online refuelling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio. This paper provides a summary of actinide transmutation schemes that have been studied in CANDU reactors at AECL, including the works performed in the past. The schemes studied include homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel. The transmutation schemes that are presented reflect several different partitioning schemes. Separation of americium, often with curium, from the other actinides enables targeted destruction of americium, which is a main contributor to the decay heat 100-1000 years after discharge from the reactor. Another scheme is group-extracted transuranic elements, in which all of the transuranic elements, plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm) are extracted together and then transmuted. This paper also addresses ways of utilizing the recycled uranium, another stream from the separation of spent nuclear fuel, in order to drive the transmutation of other actinides.

  17. Heat transfer analysis of consolidated dry storage system for CANDU spent fuel considering environmental conditions of Wolsong site

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Yoon, J. H.; Choi, B. I.; Lee, H. Y. [Korea Hydraulic and Nuclear Power Company, Taejon (Korea, Republic of)

    2004-07-01

    The purpose of the present paper is to perform heat transfer analysis of the MACSTOR/KN-400 dry storage system for CANDU spent fuel in order to predict maximum concrete temperatures and temperature gradients. This module has twice the capacity of the existing MACSTOR-200, which is in operation at Gentilly-2. In the thermal design of the MACSTOR/KN-400, Thermal Insulation Panels(TIP) were introduced to reduce concrete temperatures and temperature gradients in the module caused by the high fuel heat loads. Environmental factors such as solar heat, daily temperature variations and ambient temperatures in summer and winter at Wolsong site and the assumed presence of hot baskets were taken into consideration in the simulations. Two cases were performed for the MACSTOR/KN-400: Off-normal cases in summer and winter. The maximum local concrete temperatures were predicted to be 63 .deg. C for the off-normal case. The temperature gradients in the concrete walls and roof are predicted to be 28C and 25C for off-normal operation in summer, incorporating a 3C uncertainty. In conclusion, this paper shows that the maximum temperature for the module is expected to meet the temperature limitations of ACI 349.

  18. Seismic Structure-Soil-Structure Interaction Analysis of a Consolidated Dry Storage Module for CANDU Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Gon; Yoon, Jeong Hyoun; Kim, Sung Hwan; Yang, Ke Hyung; Lee, Heung Young; Cho, Chun Hyung [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Choi, Kyu Sup; Jeong, In Su [KONES Corporation, Seoul (Korea, Republic of)

    2005-07-01

    The MACSTOR/KN-400 module has been developed as an effective alternative to the existing stand alone concrete canister for dry storage of CANDU spent fuel. The structure is a concrete monolith of 21.67 m long and 12.66 m wide and has a height equal to 7.518 m including the bottom slab. Inside the concrete module consists of 40 storage cylinders accommodating ten 60-bundle dry storage baskets, which are suspended from the top slab and eventually restrained at 10 cm above the bottom slab with horizontal seismic restraints. The main cooling process of the MACSTOR/KN-400 module shall be by air convection through air inlets and outlets. The civil design parameters, with respect to meteorological and seismic loads applied to the module are identical to those specified for the Wolsung CANDU 3 and 4 plants, except for local site characteristics required for soilstructure interaction (SSI) analysis. It is required for the structural integrity to fulfill the licensing requirements. As per USNRC SRP Section 3.7.2, it shall be reviewed how to consider the phenomenon of coupling of the dynamic response of adjacent structures through the soil, which is referred to as structure-soil-structure interaction (SSSI). The presence of closely spaced multiple structural foundations creates coupling between the foundations of individual structures . Some observations of the actual seismic response of structures have indicated that SSSI effects do exist, but they are generally secondary for the overall structural response motions. SSSI effects, however, may be important for a relatively small structure which is to be close to a relatively large structure, while they may be generally neglected for overall structural response of a large massive structure, such as nuclear power plant. As such the scope of the present paper is to carry out a seismic SSSI analysis in case of the MACSTOR/KN- 400 module, in order to investigate whether or not SSSI effect shall be included in the overall seismic

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-05-15

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

  20. Impact of aging and material structure on CANDU plant performance

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  1. Measurement of the composition of noble-metal particles in high-burnup CANDU fuel by wavelength dispersive X-ray microanalysis

    Energy Technology Data Exchange (ETDEWEB)

    Hocking, W.H.; Szostak, F.J

    1999-09-01

    An investigation of the composition of the metallic inclusions in CANDU fuel, which contain Mo, Tc, Ru, Rh and Pd, has been conducted as a function of burnup by wavelength dispersive X-ray (WDX) microanalysis. Quantitative measurements were performed on micrometer sized particles embedded in thin sections of fuel using elemental standards and the ZAF method. Because the fission yields of the noble metals change with burnup, as a consequence of a shift from almost entirely {sup 235}U fission to mainly {sup 239}Pu fission, their inventories were calculated from the fuel power histories using the WIMS-Origin code for comparison with experiment. Contrary to expectations that the oxygen potential would be buffered by progressive Mo oxidation, little evidence was obtained for reduced incorporation of Mo in the noble-metal particles at high burnup. These surprising results are discussed with respect to the oxygen balance in irradiated CANDU fuels and the likely intrinsic and extrinsic sinks for excess oxygen. (author)

  2. Evaluation of maximum allowable temperature inside basket of dry storage module for CANDU spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Ho; Yoon, Jeong Hyoun; Chae, Kyoung Myoung; Choi, Byung Il; Lee, Heung Young; Song, Myung Jae [Nuclear Environment Technology Institute, Taejon (Korea, Republic of); Cho, Gyu Seong [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2002-10-01

    This study provides a maximum allowable fuel temperature through a preliminary evaluation of the UO{sub 2} weight gain that may occur on a failed (breached sheathing) element of a fuel bundle. Intact bundles would not be affected as the UO{sub 2} would not be in contact with the air for the fuel storage basket. The analysis is made for the MACSTOR/KN-400 to be operated in Wolsong ambient air temperature conditions. The design basis fuel is a 6-year cooled fuel bundle that, on average has reached a burnup of 7,800 MWd/MTU. The fuel bundle considered for analysis is assumed to have a high burnup of 12,000 MWd/MTU and be located in a hot basket. The MACSTOR/KN-400 has the same air circuit as the MACSTOR and the air circuit will require a slightly higher temperature difference to exit the increased heat load. The maximum temperature of a high burnup bundle stored in the new MACSTOR/KN-400 is expected to be about 9 .deg. C higher than the fuel temperature of the MACSTOR at an equivalent constant ambient temperature. This temperature increase will in turn increase the UO{sub 2} weight gain from 0.06% (MACSTOR for Wolsong conditions) to an estimated 0.13% weight gain for the MACSTOR/KN-400. Compared to an acceptable UO{sub 2} weight gain of 0.6%, we are thus expecting to maintain a very acceptable safety factor of 4 to 5 for the new module against unacceptable stresses in the fuel sheathing. For the UO{sub 2} weight gain, the maximum allowable fuel temperature was shown by 164 .deg. C.

  3. Physics and Control Assessment of AN 850 Mw(e) - Leu-Candu Reactor.

    Science.gov (United States)

    Barbone, Michelangelo

    The physics and control assessment of an 850 MW(e) Low Enriched Uranium CANDU Pressurized Heavy Water (LEU -CANDU-PHW) reactor constitute the major objective of this thesis. The use of Low Enriched Uranium fuel in the present CANDU nuclear power generating stations is recognized as economically beneficial due to reduced fuelling costs. The LEU fuel cycle is also recognized as a stepping stone to transit from the present CANDU-PHW once-through natural Uranium cycle to advanced cycles such as those based on Plutonium recycle, once-through Th + U-235 cycle, Thorium with Uranium recycle and net U-235 feed, Thorium with Uranium recycle and Plutonium feed. However, although the use of Low Enriched Uranium in the present CANDU-PHW reactor has economic advantages, and it would act as a technical bridge between the present cycle and advanced cycles, technical problems in different areas of reactor physics and fuel management were anticipated. The present thesis research work adresses the areas of reactor physics, fuel management, and control (in particular, the spatial control of large CANDU-PHW reactors). The main conclusions that have been drawn following these studies are as follows: (1) The Low Enriched Uranium Cycle is feasible in a CANDU-PHW reactor of present design and provided that: (a) The enrichment is kept relatively low (that is, about 1% instead of 0.711%); (b) the number of bundles to be replaced at every refuelling operation is about one-half that of the natural Uranium fuel case; (c) The channels are refuelled in the same direction as the coolant. (2) The response of an LEU-CANDU-PHW reactor to reactivity perturbation such as single- and two-channel refuelling operation, shim transient, shutdown-start-up transient with enrichment levels of 0.9% and 1.2% is essentially very similar {provided that certain conditions in (1) are respected} to that of the natural uranium reactor core case without any reactor reoptimization. The general behaviour of the reactor

  4. Extrapolating power-ramp performance criteria for current and advanced CANDU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Tayal, M.; Chassie, G.G

    2000-06-01

    To improve the precision and accuracy of power-ramp performance criteria for high-burnup fuel, we have examined in-reactor fuel performance data as well as out-reactor test data. The data are consistent with some of the concepts used in the current formulations for defining fuel failure thresholds, such as size of power-ramp and extent of burnup. Our review indicates that there is a need to modify some other aspects of the current formulations; therefore, a modified formulation is presented in this paper. The improvements mainly concern corrodent concentration and its relationships with threshold stress for failure. The new formulation is consistent with known and expected trends such as strength of Zircaloy in corrosive environment, timing of the release of fission products to the pellet-to-sheath gap, CANLUB coating, and fuel burnup. Because of the increased precision and accuracy, the new formulation is better able to identify operational regimes that are at risk of power-ramp failures; this predictive ability provides enhanced protection to fuel against power-ramp defects. At die same time, by removing unnecessary conservatisms in other areas, the new formulation permits a greater range of defect-free operational envelope as well as larger operating margins in regions that are, in fact, not prone to power-ramp failures. (author)

  5. Modelling of iodine-induced stress corrosion cracking in CANDU fuel

    Science.gov (United States)

    Lewis, B. J.; Thompson, W. T.; Kleczek, M. R.; Shaheen, K.; Juhas, M.; Iglesias, F. C.

    2011-01-01

    Iodine-induced stress corrosion cracking (I-SCC) is a recognized factor for fuel-element failure in the operation of nuclear reactors requiring the implementation of mitigation measures. I-SCC is believed to depend on certain factors such as iodine concentration, oxide layer type and thickness on the fuel sheath, irradiation history, metallurgical parameters related to sheath like texture and microstructure, and the mechanical properties of zirconium alloys. This work details the development of a thermodynamics and mechanistic treatment accounting for the iodine chemistry and kinetics in the fuel-to-sheath gap and its influence on I-SCC phenomena. The governing transport equations for the model are solved with a finite-element technique using the COMSOL Multiphysics® commercial software platform. Based on this analysis, this study also proposes potential remedies for I-SCC.

  6. Fission product inventory for CANFLEX-NU and CANFLEX-RU in the limiting channel of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jung, Jong Yeob

    2007-09-15

    The fission product inventory was evaluated for the 37-element fuel, CANFLEX-NU and CANFLEX-RU fuels which irradiated in the assumed limiting O6{sub m}od channel under a normal operating condition. Generally, the assessment of fission product release is divided into 3 steps which are a calculation of FGR during steady state, a calculation of a prompt release at the time of an accident, and a calculation of delayed release after the accident. Among them the first step was described in this report. The analysis was performed by using the ELESTRES code which is a computer code used in a performance evaluation of the fuel bundle during the normal irradiation. The calculation results of the fission product inventory for the 37-element fuel, CANFLEX-NU and CANFLEX-RU were compared to each other.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

  8. Comparison of MCNP4B and WIMS-AECL calculations of coolant-void-reactivity effects for uniform lattices of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    1999-05-01

    This paper compares the results of coolant-void reactivity (CVR) reactor-physics calculations performed using the Monte Carlo N-particle transport code, MCNP version 4B, with those obtained using Atomic Energy of Canada Limited's (AECL's) latest version of the Winfrith improved multigroup scheme (WIMS) code, WIMS-AECL version 2-5c. Cross sections derived from the evaluated nuclear data file version B-VI (ENDF/B-VI) are used for both the WIMS-AECL and MCNP4B calculations. The comparison is made for uniform lattices at room temperature containing either fresh natural uranium or mixed oxide (MOX) 37-element CANDU fuel. The MOX fuel composition corresponds roughly to that of irradiated CANDU fuel at a burnup of about 4500 MWd/tU. The level of agreement between the CVR predictions of WIMS-AECL and MCNP4B is studied as a function of lattice buckling (a measure of the curvature of the neutron-flux distribution) over the range from 0.0 to 4.1 m{sup -2} . For the cases studied, it is found that the absolute k values calculated by WIMS-AECL are higher than those of MCNP4B by several mk (1 mk is a change of 0.001 in k), amounts that depend on the fuel type being modelled and the particular cross-section data used. However, the agreement between WIMS-AECL and MCNP4B is much better for the CVR (i.e., the {delta}k on coolant voiding), and is relatively insensitive to the fuel type. (author)

  9. Comparison of MCNP4B and WIMS-AECL calculations of coolant-void-reactivity effects for uniform lattices of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1999-07-01

    This paper compares the results of coolant-void reactivity (CVR) reactor-physics calculations performed using the Monte Carlo N-particle transport code, MCNP version 4B, with those obtained using Atomic Energy of Canada Limited's (AECL's) latest version of the Winfrith improved multigroup scheme (WIMS) code, WIMS-AECL version 2-5c. Cross sections derived from the evaluated nuclear data file version B-VT (ENDF/B-VI) are used for both the WIMS-AECL and MCNP4B calculations. The comparison is made for uniform lattices at room temperature containing either fresh natural uranium or mixed oxide (MOX) 37-element CANDU fuel. The MOX fuel composition corresponds roughly to that of irradiated CANDU fuel at a burnup of about 4500 MWd/tU. The level of agreement between the CVR predictions of WIMS-AECL and MCNP4B is studied as a function of lattice buckling (a measure of the curvature of the neutron-flux distribution) over the range from 0.0 to 4.1 m{sup -2}. For the cases studied, it is found that the absolute keff values calculated by WIMS-AECL are higher than those of MCNP4B by several mk (1 mk is a change of 0.001 in keff), amounts that depend on the fuel type being modelled and the particular cross-section data used. However, the agreement between WIMS-AECL and MCNP4B is much better for the CVR (i.e., the {delta}keff on coolant voiding), and is relatively insensitive to the fuel type. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-15

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

  11. Validation of physics and thermalhydraulic computer codes for advanced Candu reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Wren, D.J.; Popov, N.; Snell, V.G. [Atomic Energy of Canada Ltd, (Canada)

    2004-07-01

    Atomic Energy of Canada Ltd. (AECL) is developing an Advanced Candu Reactor (ACR) that is an evolutionary advancement of the currently operating Candu 6 reactors. The ACR is being designed to produce electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular Candu concept of horizontal fuel channels surrounded by a heavy water moderator. However, ACR uses slightly enriched uranium fuel compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (via large reductions in the heavy water moderator volume and replacement of the heavy water coolant with light water coolant) and improved safety. AECL has developed and implemented a software quality assurance program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. Since the basic design of the ACR is equivalent to that of the Candu 6, most of the key phenomena associated with the safety analyses of ACR are common, and the Candu industry standard tool-set of safety analysis codes can be applied to the analysis of the ACR. A systematic assessment of computer code applicability addressing the unique features of the ACR design was performed covering the important aspects of the computer code structure, models, constitutive correlations, and validation database. Arising from this assessment, limited additional requirements for code modifications and extensions to the validation databases have been identified. This paper provides an outline of the AECL software quality assurance program process for the validation of computer codes used to perform physics and thermal-hydraulics safety analyses of the ACR. It describes the additional validation work that has been identified for these codes and the planned, and ongoing, experimental programs to extend the code validation as required to address specific ACR design

  12. Validation of WIMS-CANDU using Pin-Cell Lattices

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won Young; Min, Byung Joo; Park, Joo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

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

  13. Mirrored serpentine flow channels for fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Rock, Jeffrey Allan (Rochester, NY)

    2000-08-08

    A PEM fuel cell having serpentine flow field channels wherein the input/inlet legs of each channel border the input/inlet legs of the next adjacent channels in the same flow field, and the output/exit legs of each channel border the output/exit legs of the next adjacent channels in the same flow field. The serpentine fuel flow channels may be longer, and may contain more medial legs, than the serpentine oxidant flow channels.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  15. Input modelling for subchannel analysis of CANFLEX fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-01

    This report describs the input modelling for subchannel analysis of CANFLEX fuel bundle using CASS(Candu thermalhydraulic Analysis by Subchannel approacheS) code which has been developed for subchannel analysis of CANDU fuel channel. CASS code can give the different calculation results according to users' input modelling. Hence, the objective of this report provide the background information of input modelling, the accuracy of input data and gives the confidence of calculation results. (author). 11 refs., 3 figs., 4 tabs.

  16. Measurements of grain-boundary inventories of {sup 137}Cs, {sup 90}Sr and {sup 99}Tc in used CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Stroes-Gascoyne, S.; Tait, J.C.; Porth, R.J.; McConnell, J.L.; Barnsdale, T.R.; Watson, S. [Whiteshell Labs., Pinawa, Manitoba (Canada)

    1993-12-31

    Two methods were used to measure grain-boundary inventories of {sup 137}Cs, {sup 90}Sr and {sup 99}Tc in used CANDU fuel, to corroborate source term estimates based on a fission gas release code. Used fuels were partially oxidized at 200{degrees}C in air to overall compositions of UO{sub 2+x} (0.15{<=} {times} {<=}0.25) to expose UO{sub 2} grain boundaries, followed by leaching in aqueous solution. Only a fraction (2 to 18%) of the calculated gap + grain-boundary inventories for {sup 137}Cs was released. This suggests that the calculations overestimate Cs release or that oxidation does not expose all grain boundaries, or that Cs release from grain boundaries is slow. Release of {sup 90}Sr (0.01 to 0.7%) agreed reasonably well with the source term estimates (0.001 to 0.3%). Release of {sup 99}Tc (0.3 to 1.5%) suggests that the source term estimate for the upper involved leaching of crushed and side-fractionated used fuel in either a static or dynamic system. A direct one-to-one correlation between calculated and measured gap + grain-boundary inventories for {sup 137}Cs was found for low- and medium-power fuels.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  18. Assessment of CANDU reactor physics effects using a simplified whole-core MCNP model

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    2002-07-01

    A whole-core Monte Carlo n-particle (MCNP) model of a simplified CANDU reactor was developed and used to study core configurations and reactor physics phenomena of interest in CANDU safety analysis. The resulting reactivity data were compared with values derived from corresponding WIMS-AECL/RFSP, two-neutron-energy-group diffusion theory core simulations, thereby extending the range of CANDU-related code-to-code benchmark comparisons to include whole-core representations. These comparisons show a systematic discrepancy of about 6 mk between the respective absolute k{sub eff} values, but very good agreement to within about -0.15 {+-} 0.06 mk for the reactivity perturbation induced by G-core checkerboard coolant voiding. These findings are generally consistent with the results of much simpler uniform-lattice comparisons involving only WIMS-AECL and MCNP. In addition, MCNP fission-energy tallies were used to evaluate other core-wide properties, such as fuel bundle and total-channel power distributions, as well as intra-bundle details, such as outer-fuel-ring relative power densities and outer-ring fuel element azimuthal power variations, which cannot be determined directly from WIMS-AECL/RFSP core calculations. The average MCNP values for the ratio of outer fuel element to average fuel element power density agreed well with corresponding values derived from WIMS-AECL lattice-cell cases, showing a small systematic discrepancy of about 0.5 %, independent of fuel bum-up. For fuel bundles containing the highest-power fuel elements, the maximum peak-to-average outer-element azimuthal power variation was about 2.5% for cases where a statistically significant trend was observed, while much larger peak-to-average outer-element azimuthal power variations of up to around 42% were observed in low-power fuel bundles at the core/radial-neutron-reflector interface. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  20. CANDU, building the future

    Energy Technology Data Exchange (ETDEWEB)

    Stern, F. [Stern Laboratories (Canada)

    1997-07-01

    The CEO of Stern Laboratories delivered a speech on the problems and challenges facing the nuclear industry. The CANDU system is looked at as the practical choice for the future of our energy source. The people of the industry must be utilized and respected to deliver to the best of their ability.

  1. Effect of DUPIC cycle on CANDU reactor safety parameters

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  2. Plutonium Consumption Program, CANDU Reactor Project final report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-31

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

  3. Validation of WIMS-AECL reactivity device calculations for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Min, Byung Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Donnelly, J. V. [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1997-06-01

    An important component of the overall program to validate WIMS-AECL for use with RFSP in the analysis of CANDU-6 reactors for design and safety analysis calculations is the validation of calculations of incremental cross sections used to represent reactivity devices. A method has been developed for the calculation of the three-dimensional neutron flux distribution in and around CANDU reactor fuel channels and reactivity control devices. The methods is based on one- and two dimensional transport calculations with the WIMS-AECL lattice cell code, SPH homogenization, and three-dimensional flux calculations with finite-difference diffusion theory using the MULTICELL code. Simulations of Wolsung 1 Phase-B commissioning measurements and Point Lepreau restart tests have been performed, as a part of the program to validate WIMS-AECL lattice cell calculations for application to CANDU reactor simulations in RFSP. The incremental cross section properties of the Wolsung 1 and Point Lepreau adjusters, Mechanical Control Absorbers(MCA) and liquid Zone Control Unit (ZCU) is based on the WIMS-AECL/MULTICELL modelling methods and the results are compared with those of WIMS-AECL/DRAGON-2 modelling methods. (author). 13 tabs., 4 figs., 11 refs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  5. Neutronics-thermalhydraulics coupling in a CANDU SCWR

    Science.gov (United States)

    Adouki, Pierre

    In order to implement new nuclear technologies as a solution to the growing demand for energy, 10 countries agreed on a framework for international cooperation in 2002, to form the Generation IV International Forum (GIF). The goal of the GIF is to design the next generation of nuclear reactors that would be cost effective and would enhance safety. This forum has proposed several types of Generation IV reactors including the Supercritical Water-Cooled Reactor (SCWR). The SCWR comes in two main configurations: pressure vessel SCWR and pressure tube SCWR (PT-SCWR). In this study, the CANDU SCWR (a PT-SCWR) is considered. This reactor is oriented vertically and contains 336 channels with a length of 5 m. The target coolant inlet and outlet temperatures are 350 Celsius and 625 Celsius, respectively. The coolant flows downwards, and the reactor power is 2540 MWth. Various fuel designs have been considered in order not to exceed the linear element rating. However, the dependency between the core power and thermalhydraulics parameters results in the necessity to use a neutronics/thermalhydaulics coupling scheme to determine the core power and the thermalhydraulics parameters. The core power obtained has a power peaking factor of 1.4. The bundle power distribution for all channels has a peak at the third bundle from the inlet, but the value of this peak increases with the channel power. The heat-transfer coefficient and the specific-heat capacity have a peak at the same location in a channel, and this location shifts toward the inlet as the channel power increases. The exit coolant temperature increases with the channel power, while the exit coolant density and pressure decrease with the channel power. Also, higher channel powers lead to higher fuel and cladding temperatures. Moreover, as the coupling method is applied, the effective multiplication factor and the values of thermalhydaulics parameters oscillate as they converge.

  6. Estimation of Aging Effects on LOHS for CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Yong Ki; Moon, Bok Ja; Kim, Seoung Rae [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    To evaluate the Wolsong Unit 1's capacity to respond to large-scale natural disaster exceeding design, the loss of heat sink(LOHS) accident accompanied by loss of all electric power is simulated as a beyond design basis accident. This analysis is considered the aging effects of plant as the consequences of LOHS accident. Various components of primary heat transport system(PHTS) get aged and some of the important aging effects of CANDU reactor are pressure tube(PT) diametral creep, steam generator(SG) U-tube fouling, increased feeder roughness, and feeder orifice degradation. These effects result in higher inlet header temperatures, reduced flows in some fuel channels, and higher void fraction in fuel channel outlets. Fresh and aged models are established for the analysis where fresh model is the circuit model simulating the conditions at retubing and aged model corresponds to the model reflecting the aged condition at 11 EFPY after retubing. CATHENA computer code[1] is used for the analysis of the system behavior under LOHS condition. The LOHS accident is analyzed for fresh and aged models using CATHENA thermal hydraulic computer code. The decay heat removal is one of the most important factors for mitigation of this accident. The major aging effect on decay heat removal is the reduction of heat transfer efficiency by steam generator. Thus, the channel failure time cannot be conservatively estimated if aged model is applied for the analysis of this accident.

  7. Development of the advanced CANDU technology

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Soo Dong; Min, Byung Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Na, Y. H.; Lee, S. Y.; Choi, J. H.; Lee, B. C.; Kim, S. N.; Jo, C. H.; Paik, J. S.; On, M. R.; Park, H. S.; Kim, S. R. [Korea Electric Power Co., Taejon (Korea, Republic of)

    1997-07-01

    The purpose of this study is to develop the advanced design technology to improve safety, operability and economy and to develop and advanced safety evaluation system. More realistic and reasonable methodology and modeling was employed to improve safety margin in containment analysis. Various efforts have been made to verify the CATHENA code which is the major safety analysis code for CANDU PHWR system. Fully computerized prototype ECCS was developed. The feasibility study and conceptual design of the distributed digital control system have been performed as well. The core characteristics of advanced fuel cycle, fuel management and power upgrade have been studied to determine the advanced core. (author). 77 refs., 51 tabs., 108 figs.

  8. A neutronic study of the cycle PWR-CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alberto da; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Fortini, Angela; Pinheiro, Ricardo Brant [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear]. E-mail: albertomoc@terra.com.br; claubia@nuclear.ufmg.br; dora@nuclear.ufmg.br; fortini@nuclear.ufmg.br; rbp@nuclear.ufmg.br

    2007-07-01

    The cycle PWR-CANDU was simulated using the WIMSD-5B and ORIGEN2.1 codes. It was simulated a fuel burnup of 33,000 MWd/t for UO{sub 2} with enrichment of 3.2% and a fuel extended burnup of 45,000 MWd/t for UO{sub 2} with enrichments of 3.5%, 4.0% and 5.0% in a PWR reactor. The PWR discharged fuel was submitted to the simulation of deposition for five years. After that, it was submitted to AYROX reprocessing and used to produce a fuel to CANDU reactor. Then, it was simulated the burnup in the CANDU. Parameters such as infinite medium multiplication factor, k{sub inf}, fuel temperature coefficient of reactivity, {alpha}{sub TF}, moderator temperature coefficient of reactivity, {alpha}{sub TM}, the ratio rapid flux/total flux and the isotopic composition in the begin and the end of life were evaluated. The results showed that the fuels analyzed could be used on PWR and CANDU reactors without the need of change on the design of these reactors. (author)

  9. Analyse du transfert de chaleur et de la perte de pression pour des ecoulements supercritiques dans le reacteur CANDU-SCWR

    Science.gov (United States)

    Zoghlami, Sarra

    The supercritical water reactor is one of the six concepts of generation IV nuclear reactors that has been selected by the International Generation IV Forum (GIF). Canada has chosen to conduct advanced research on this type of reactor. For the design and safety analysis of the reactor concept, the development of numerical simulation codes is needed. The ARTHUR code is a thermal-hydraulic computer code developed by Fassi-Fehri (2008), at the Ecole Polytechnique de Montreal, to analyse the CANDU-6 reactor. The purpose of this project is to modify this numerical code so that it can be used to treat the CANDU-SCWR. To calculate the coolant thermal-hydraulics properties in the fuel channel of a CANDU-SCWR, it was assumed that the water flows under supercritical conditions is a one-phase flow. Thus within this code, we developed the conservation equations for one-phase flow. Hydraulic resistance and heat transfer at supercritical pressure are two important aspects to be considered in the modeling of a fuel channel in a nuclear reactor. To choose the accurate correlation to predict the pressure friction factor, we compared numerical calculations, using different correlations found in literature, to experimental data. We concluded that the Garimella (2008) correlation is the most consistent, to be incorporated in the ARTHUR &barbelow;SCWR code. We proved that the choice of the friction factor correlation affects slightly the distribution of thermal-hydraulic properties in the fuel channel. Under supercritical conditions, water thermal-physical properties are characterized by significant variations in the pseudo-critical region. This behavior influences the forced convection heat transfer phenomena. To choose the adequate correlation to calculate the forced convection heat transfer coefficient, we compared numerical results to experimental data, and we found that the standard deviation given by Mokry et al. (2010) correlation is the lowest. In order to model the fuel

  10. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2000-03-15

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model if existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA analysis. There are three main area of model development, i.e. moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version.

  11. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA Analysis. There are three main area of model development, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version. 15 refs., 37 figs., 8 tabs. (Author)

  12. Large-break loss-of-coolant accident phenomena identification and ranking table (PIRT) for the advanced Candu reactor

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.; Snell, V.G.; Sills, H.E.; Langman, V.J.; Boyack, B. [Atomic Energy of Canada Ltd (Canada)

    2004-07-01

    The Advanced Candu Reactor (ACR) is an evolutionary advancement of the current Candu-6 reactor, aimed at producing electrical power for a capital cost and unit-energy cost significantly less than that of current reactor designs. The ACR retains the modular concept of horizontal fuel channels surrounded by heavy water moderator, as with all Candu reactors. However, ACR uses slightly enriched uranium (SEU) fuel, compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (e.g., via reductions in the heavy water requirements and the use of a light water coolant), as well as improved safety. This paper is focused on the double-ended guillotine critical inlet header break (CRIHB) loss-of-coolant accident (LOCA) in an ACR reactor, which is considered as a large break LOCA. Large Break LOCA in water-cooled reactors has been used historically as a design basis event by regulators, and it has attracted a very large share of safety analysis and regulatory review. The LBLOCA event covers a wide range of system behaviours and fundamental phenomena. The Phenomena Identification and Ranking Table (PIRT) for LBLOCA therefore provides a good understanding of many of the safety characteristics of the ACR design. The paper outlines the design characteristics of the ACR reactor that impact the PIRT process and computer code applicability. It also describes the LOCA phenomena, lists all components and systems that have an important role during the event, discusses the PIRT process and results, and presents the final PIRT summary table. (authors)

  13. A study on the direct use of spent PWR fuel in CANDU reactors -Development of DUPIC fuel on manufacturing and quality control technology-

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Park, Hyun Soo; Lee, Yung Woo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    Oxidation/reduction process was established after analysis of the effect of process parameter on the sintering behavior using SIMFUEL. Process equipment was studied more detail and some of process equipment items were designed and procured. The chemical analysing method of fission products and fissile content in DUPIC fuel was studied and the behavior and the characteristics of fission products in fuel was also done. Requirement for irradiation in HANARO was analysed to prepare performance evaluation. 100 figs, 48 tabs, 170 refs. (Author).

  14. Sensitivity of Candu-Scwr Reactor Physics Calculations to Nuclear Data Files

    Science.gov (United States)

    Kozier, K. S.; Dyck, G. R.

    2006-04-01

    A simplified MCNP model of a CANDU-SCWR lattice was used to test the sensitivity of the calculated reactivity to various nuclear data files involving issues of current interest. These tests were performed for cooled and voided conditions, with and without axial neutron leakage, for a uniform lattice of mid-life fuel and a mixed lattice with high-burnup and low-burnup fuel in alternate channels. Results were compared using different room-temperature data files for deuterium, various thermal-scattering-law data files for hydrogen bound in light water and deuterium bound in heavy water, and for pre-ENDF/B-VII and ENDF/B-VI.8 data for uranium. The reactivity differences observed were small (typically <1 mk) and increased with axial neutron leakage.

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

    Science.gov (United States)

    Rochefort, P. A.

    2009-02-01

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

  16. Probabilistic fracture mechanics applied for DHC assessment in the cool-down transients for CANDU pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Radu, Vasile, E-mail: vasile.radu@nuclear.ro [Institute for Nuclear Research Pitesti, 1st Campului Street, 115400 Mioveni, Arges, P.O. Box 78, Mioveni (Romania); Roth, Maria [Institute for Nuclear Research Pitesti, 1st Campului Street, 115400 Mioveni, Arges, P.O. Box 78, Mioveni (Romania)

    2012-12-15

    For CANDU pressure tubes made from Zr-2.5%Nb alloy, the mechanism called delayed hydride cracking (DHC) is widely recognized as main mechanism responsible for crack initiation and propagation in the pipe wall. Generation of some blunt flaws at the inner pressure tube surface during refueling by fuel bundle bearing pad or by debris fretting, combined with hydrogen/deuterium up-take (20-40 ppm) from normal corrosion process with coolant, may lead to crack initiation and growth. The process is governed by hydrogen hysteresis of terminal solid solubility limits in Zirconium and the diffusion of hydrogen atoms in the stress gradient near to a stress spot (flaw). Creep and irradiation growth under normal operating conditions promote the specific mechanisms for Zirconium alloys, which result in circumferential expansion, accompanied by wall thinning and length increasing. These complicate damage mechanisms in the case of CANDU pressure tubes that are also are affected by irradiation environment in the reactor core. The structural integrity assessment of CANDU fuel channels is based on the technical requirements and methodology stated in the Canadian Standard N285.8. Usually it works with fracture mechanics principles in a deterministic manner. However, there are inherent uncertainties from the in-service inspection, which are associated with those from material properties determination; therefore a necessary conservatism in deterministic evaluation should be used. Probabilistic approach, based on fracture mechanics principle and appropriate limit state functions defined as fracture criteria, appears as a promising complementary way to evaluate structural integrity of CANDU pressure tubes. To perform this, one has to account for the uncertainties that are associated with the main parameters for pressure tube assessment, such as: flaws distribution and sizing, initial hydrogen concentration, fracture toughness, DHC rate and dimensional changes induced by long term

  17. Basic research and industrialization of CANDU advanced fuel - Effect of transverse convex curvature on boiling heat transfer and ONB point of nucleate fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Chun; Lee, Young; Lee, Sung Hong [Pusan National University, Pusan (Korea)

    2000-04-01

    Recently, the effect of convex curvature on heat transfer should not be ignored when the radius of curvature tends to be small and/or associated with high heat transfer rate cases. Both analytical and experimental studies were performed to prove the effect of transverse convex curvature on the boiling heat transfer in concentric annuli flows. The effect of the transverse convex surface curvature on ONB are studied analytically in the case of reactor and evaporator. It is seen that the inner wall heat flux depends on R/sub i/, Rc, Re, Pr, {alpha}, and the {theta} of working fluid. An experimental study on the incipience of nucleate boiling is performed as a verification ad extension of previous analyses. Through flow visualization, the results show that the most dominant parameter to affect the heat flux at ONB is found to be the surface curvature. The heat flux data at ONB increases with the Re and the subcooling, and the effect of subcooling on ONB becomes smaller with decreasing Re. The heat flux at ONB increases rapidly as increase in {alpha} due to higher convective motion of bulk flow. Comparison between both results are accomplished with respect to the relative enhancement due to the convex curvature. The relative heat transfer enhancement ratio shows a good agreement between theory and experiment qualitatively and quantitatively. In conclusion, the obtained results suggest that the effect transverse convex curvature appears significantly in the boiling heat transfer. Therefore, it can be clearly expected that the effect should be more strong at the case of critical heat flux condition which is the most important design goal of the advanced nuclear fuel rods. 30 refs., 78 figs. (Author)

  18. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR PART 4 - FUEL CHANNEL ASSEMBLY

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2014-05-01

    Full Text Available As many nuclear power plants are reaching their end of lifecycle, the decommissioning of these installations has become one of the 21st century’s great challenges. Each project may be managed differently, depending on the country, development policies, financial considerations, and the availability of qualified engineers or specialized companies to handle such projects. The principle objective of decommissioning is to place a facility into such a condition that there is no unacceptable risk from the decommissioned facility to public health and safety of the environment. In order to ensure that at the end of its life the risk from a facility is within acceptable bounds, action is normally required. The overall decommissioning strategy is to deliver a timely, costeffective program while maintaining high standards of safety, security and environmental protection. If facilities were not decommissioned, they could degrade and potentially present an environmental radiological hazard in the future. Simply abandoning or leaving a facility after ceasing operations is not considered to be an acceptable alternative to decommissioning. The final aim of decommissioning is to recover the geographic site to its original condition.

  19. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR PART 3 - FUEL CHANNEL REFERENCES

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2014-05-01

    Full Text Available As many nuclear power plants are reaching their end of lifecycle, the decommissioning of these installations has become one of the 21st century’s great challenges. Each project may be managed differently, depending on the country, development policies, financial considerations, and the availability of qualified engineers or specialized companies to handle such projects. The principle objective of decommissioning is to place a facility into such a condition that there is no unacceptable risk from the decommissioned facility to public health and safety of the environment. In order to ensure that at the end of its life the risk from a facility is within acceptable bounds, action is normally required. The overall decommissioning strategy is to deliver a timely, cost-effective program while maintaining high standards of safety, security and environmental protection. If facilities were not decommissioned, they could degrade and potentially present an environmental radiological hazard in the future. Simply abandoning or leaving a facility after ceasing operations is not considered to be an acceptable alternative to decommissioning. The final aim of decommissioning is to recover the geographic site to its original condition.

  20. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR PART 2 - FUEL CHANNEL PRESENTATION

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2014-05-01

    Full Text Available As many nuclear power plants are reaching their end of lifecycle, the decommissioning of these installations has become one of the 21st century’s great challenges. Each project may be managed differently, depending on the country, development policies, financial considerations, and the availability of qualified engineers or specialized companies to handle such projects. The principle objective of decommissioning is to place a facility into such a condition that there is no unacceptable risk from the decommissioned facility to public health and safety of the environment. In order to ensure that at the end of its life the risk from a facility is within acceptable bounds, action is normally required. The overall decommissioning strategy is to deliver a timely, costeffective program while maintaining high standards of safety, security and environmental protection. If facilities were not decommissioned, they could degrade and potentially present an environmental radiological hazard in the future. Simply abandoning or leaving a facility after ceasing operations is not considered to be an acceptable alternative to decommissioning.The final aim of decommissioning is to recover the geographic site to its original condition.

  1. The DUPIC fuel development program in KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Yang, M. S.; Park, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-07-01

    This study describes the DUPIC fuel development program in KAERI as follows; Burning spent PWR fuel again in CANDU by DUPIC, Compatibility with existing CANDU system, Feasibility of DUPIC fuel fabrication, Waste reduction, Safeguard ability, Economics of DUPIC fuel cycle, The DUPIC fuel development program, and International prospective. 5 refs., 10 figs.

  2. Technology channel fuel cells; Reseau technologique piles a combustible

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    This document presents the PACo channel, its research and development program and the calendar of the first year. The PACo channel aims at stimulate the technology innovation in the domain of the fuel cells and organize collaborations between enterprises and research laboratories. (A.L.B.)

  3. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, M. K.; Lee, W. J. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-03-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a fourth step of the whole project, applying the RELAP5/MOD3/CANDU+ version for the real CANDU plant LOCA Analysis and D2O leakage incident. There are three main models under investigation, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs, especially when CANDU LOCA is tested. Also, for Wolsung unit 1 D2O leakage incident analysis, the plant behavior is predicted with the newly developed version for the first 1000 seconds after onset of the incident, with the main interest aiming for system pressure, level control system, and thermal hydraulic transient behavior of the secondary system. The model applied for this particular application includes heat transfer model of nuclear fuel assembly, decay heat model, and MOV (Motor Operated Valve) model. Finally, the code maintenance work, mainly correcting the known errors, is presented. 12 refs., 26 figs., 3 tabs. (Author)

  4. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THECANDU 6 NUCLEAR REACTOR. PART 8 - PRESENTATION OF THE CUTTING AND EXTRACTING DEVICE

    Directory of Open Access Journals (Sweden)

    Constantin POPESCU

    2015-05-01

    Full Text Available This paper present a constructive solution proposed by the authors in order to achieve of a cutting and extracting device for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. One of the most important part of the decommissioning device is the Cutting and Extraction Device (CED which perform the dismantling, cutting and extraction of the fuel channel components. This flexible and modular device is designed to work inside the fuel channel. The main operations performed by the Cutting and Extraction Device (CED are dismantling and extraction of the channel closure plug and shield plug, cutting and extraction of the pressure tube. The Cutting and Extraction Device (CED consists of following modules: guiding-fixing module, traction modules, cutting module, guiding-extracting module and articulated elements for modules connecting. The guiding-fixing module is equipped with elastic guiding rollers and fixing claws in working position, the traction modules are provided with variable pitch rollers for allowing travel speed change through the fuel channel. The cutting module is positioned in the middle of the device and it is equipped with three roll knives for pressure tube cutting, having a system for cutting place video surveillance and pyrometers for cutting place monitoring temperature. The operations performed by the Cutting and Extraction Device (CED of fuel channel are as follows: unblock and extract the channel closure plug, unblock and extract the channel shield plug, block and cut the middle of the pressure tube, block and cut the end of the pressure tube, block and extract the half of pressure tube. The Cutting and Extraction Device (CED is fully automated, connected by wires to a Programmable Logic Controller (PLC and controlled from a Human Machine Interface (HMI. The design of the Cutting and Extraction Device (CED shall be achieved according to the particular features of the fuel channel components to be

  5. Thermalhydraulics of advanced 37-element fuel bundle in crept pressure tubes

    Directory of Open Access Journals (Sweden)

    Park Joo Hwan

    2016-01-01

    Full Text Available A CANDU-6 reactor, which has 380 fuel channels of a pressure tube type, is suffering from aging or creep of the pressure tubes. Most of the aging effects for the CANDU primary heat transport system were originated from the horizontal crept pressure tubes. As the operating years of a CANDU reactor proceed, a pressure tube experiences high neutron irradiation damage under high temperature and pressure. The crept pressure tube can deteriorate the Critical Heat Flux (CHF of a fuel channel and finally worsen the reactor operating performance and thermal margin. Recently, the modification of the central subchannel area with increasing inner pitch length of a standard 37-element fuel bundle was proposed and studied in terms of the dryout power enhancement for the uncrept pressure tube since a standard 37-element fuel bundle has a relatively small flow area and high flow resistance at the central region. This study introduced a subchannel analysis for the crept pressure tubes loaded with the inner pitch length modification of a standard 37-element fuel bundle. In addition, the subchannel characteristics were investigated according to the flow area change of the center subchannels for the crept pressure tubes. Also, it was discussed how much the crept pressure tubes affected the thermalhydraulic characteristics of the fuel channel as well as the dryout power for the modification of a standard 37-element fuel bundle.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  7. Evaluation of CANDU6 PCR (power coefficient of reactivity) with a 3-D whole-core Monte Carlo Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Motalab, Mohammad Abdul; Kim, Woosong; Kim, Yonghee, E-mail: yongheekim@kaist.ac.kr

    2015-12-15

    Highlights: • The PCR of the CANDU6 reactor is slightly negative at low power, e.g. <80% P. • Doppler broadening of scattering resonances improves noticeably the FTC and make the PCR more negative or less positive in CANDU6. • The elevated inlet coolant condition can worsen significantly the PCR of CANDU6. • Improved design tools are needed for the safety evaluation of CANDU6 reactor. - Abstract: The power coefficient of reactivity (PCR) is a very important parameter for inherent safety and stability of nuclear reactors. The combined effect of a relatively less negative fuel temperature coefficient and a positive coolant temperature coefficient make the CANDU6 (CANada Deuterium Uranium) PCR very close to zero. In the original CANDU6 design, the PCR was calculated to be clearly negative. However, the latest physics design tools predict that the PCR is slightly positive for a wide operational range of reactor power. It is upon this contradictory observation that the CANDU6 PCR is re-evaluated in this work. In our previous study, the CANDU6 PCR was evaluated through a standard lattice analysis at mid-burnup and was found to be negative at low power. In this paper, the study was extended to a detailed 3-D CANDU6 whole-core model using the Monte Carlo code Serpent2. The Doppler broadening rejection correction (DBRC) method was implemented in the Serpent2 code in order to take into account thermal motion of the heavy uranium nucleus in the neutron-U scattering reactions. Time-average equilibrium core was considered for the evaluation of the representative PCR of CANDU6. Two thermal hydraulic models were considered in this work: one at design condition and the other at operating condition. Bundle-wise distributions of the coolant properties are modeled and the bundle-wise fuel temperature is also considered in this study. The evaluated nuclear data library ENDF/B-VII.0 was used throughout this Serpent2 evaluation. In these Monte Carlo calculations, a large number

  8. Development of modern CANDU PHWR cross-section libraries for SCALE

    Energy Technology Data Exchange (ETDEWEB)

    Shoman, Nathan T., E-mail: nshoman@vols.utk.edu; Skutnik, Steven E., E-mail: sskutnik@utk.edu

    2016-06-15

    Highlights: • New ORIGEN libraries for CANDU 28 and 37-element fuel assemblies have been created. • These new reactor data libraries are based on modern ENDF/B-VII.0 cross-section data. • The updated CANDU data libraries show good agreement with radiochemical assay data. • Eu-154 overestimated when using ENDF-VII.0 due to a lower thermal capture cross-section. - Abstract: A new set of SCALE fuel lattice models have been developed for the 28-element and 37-element CANDU fuel assembly designs using modern cross-section data from ENDF-B/VII.0 in order to produce new reactor data libraries for SCALE/ORIGEN depletion analyses. These new libraries are intended to provide users with a convenient means of evaluating depletion of CANDU fuel assemblies using ORIGEN through pre-generated cross sections based on SCALE lattice physics calculations. The performance of the new CANDU ORIGEN libraries in depletion analysis benchmarks to radiochemical assay data were compared to the previous version of the CANDU libraries provided with SCALE (based on WIMS-AECL models). Benchmark comparisons with available radiochemical assay data indicate that the new cross-section libraries perform well at matching major actinide species (U/Pu), which are generally within 1–4% of experimental values. The library also showed similar or better results over the WIMS-AECL library regarding fission product species and minor actinoids (Np, Am, and Cm). However, a notable exception was in calculated inventories of {sup 154}Eu and {sup 155}Eu, where the new library employing modern nuclear data (ENDF/B-VII.0) performed substantially poorer than the previous WIMS-AECL library (which used ENDF-B/VI.8 cross-sections for these species). The cause for this discrepancy appears to be due to differences in the {sup 154}Eu thermal capture cross-section between ENDF/B-VI.8 and ENDF/B-VII.0, an effect which is exacerbated by the highly thermalized flux of a CANDU heavy water reactor compared to that of a

  9. Some applications related to the structural integrity analyses of CANDU 6 pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Radu, V.S. [Inst. for Nuclear Research Pitesti, Pitesti (Romania)]. E-mail: vradu@nuclear.ro

    2006-07-01

    The flaws found during in-service inspection of Zr-2.5%Nb CANDU pressure tubes include fuel bundle bearing pad fretting flaws and debrise fretting flaws. In-service flaws are evaluated using fitness-for-service procedures to justify continued operation of pressure tube containing the flaw. The flaw evaluation procedures address crack initiation due to Delayed Hydride Cracking (DHC) under constant loading and also address fracture initiation and plastic collapse. The paper presents some applications related to the influence of the residual hoop stresses at roll-expanded joint region into stainless steel fittings at both ends on the structural integrity evaluation in the presence of blunt flaws. Two cases of blunt flaws were considered for evaluation: fuel bundle bearing pad fretting flaws and debrise fretting flaws. The blunt flaw geometry modeling and stress-strain analyses were performed by means finite element method (FEM) with FEA-Flaw computer code. The allowable peak flaw-tip stress and the Failure Assessment Diagram (FAD) for DHC initiation criterion were used for integrity assessment for the mentioned blunt flaws. Applications are performed as part of the research program address to evaluation of the in-service inspection results of the fuel channels from Cernavoda NPP. (author)

  10. LBB in Candu plants

    Energy Technology Data Exchange (ETDEWEB)

    Kozluk, M.J.; Vijay, D.K. [Ontario Hydro Nuclear, Toronto, Ontario (Canada)

    1997-04-01

    Postulated catastrophic rupture of high-energy piping systems is the fundamental criterion used for the safety design basis of both light and heavy water nuclear generating stations. Historically, the criterion has been applied by assuming a nonmechanistic instantaneous double-ended guillotine rupture of the largest diameter pipes inside of containment. Nonmechanistic, meaning that the assumption of an instantaneous guillotine rupture has not been based on stresses in the pipe, failure mechanisms, toughness of the piping material, nor the dynamics of the ruptured pipe ends as they separate. This postulated instantaneous double-ended guillotine rupture of a pipe was a convenient simplifying assumption that resulted in a conservative accident scenario. This conservative accident scenario has now become entrenched as the design basis accident for: containment design, shutdown system design, emergency fuel cooling systems design, and to establish environmental qualification temperature and pressure conditions. The requirement to address dynamic effects associated with the postulated pipe rupture subsequently evolved. The dynamic effects include: potential missiles, pipe whipping, blowdown jets, and thermal-hydraulic transients. Recent advances in fracture mechanics research have demonstrated that certain pipes under specific conditions cannot crack in ways that result in an instantaneous guillotine rupture. Canadian utilities are now using mechanistic fracture mechanics and leak-before-break assessments on a case-by-case basis, in limited applications, to support licensing cases which seek exemption from the need to consider the various dynamic effects associated with postulated instantaneous catastrophic rupture of high-energy piping systems inside and outside of containment.

  11. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THECANDU 6 NUCLEAR REACTOR. PART 11 - PRESENTATION OF THE CUTTING AND EXTRACTING DEVICE OPERATING

    Directory of Open Access Journals (Sweden)

    Constantin D. STANESCU

    2015-05-01

    Full Text Available This paper presents a constructive solution proposed by the authors in order to achieve of a cutting and extracting device for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The Cutting and Extraction Device (CED performs the dismantling, cutting and extraction of the fuel channel components. It's a flexible and modular device, which is designed to work inside the fuel channel and has the following functions: moving with variable speed, temperature monitoring and video surveillance inside the pipe, unblock and extract the channel closure plug (from End Fitting - EF, unblock and extract the channel shield plug (from Lattice Tube - LT, block and cut the middle of the pressure tube, block and cut the end of the pressure tube, block and extract the half of pressure tube. The Cutting and Extraction Device (CED consists of following modules: guiding-fixing module, traction modules, cutting module, guiding-extracting module and flexible elements for modules connecting. The guiding-fixing module is equipped with elastic guiding rollers and fixing claws in working position, the traction modules are provided with variable pitch rollers for allowing variable travel speed through the fuel channel. The cutting module is positioned in the middle of the device and it is equipped with three knife rolls for pressure tube cutting, using a system for cutting place video surveillance and pyrometers for monitoring cutting place temperature. The Cutting and Extraction Device (CED is fully automated, connected by wires to a Programmable Logic Controller (PLC and controlled from a Human Machine Interface (HMI. The design of the Cutting and Extraction Device (CED shall be achieved according to the particular features of the fuel channel components to be dismantled and to ensure radiation protection of workers.

  12. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR PART 5 - FUEL CHANEL DECOMMISSIONING

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2014-05-01

    Full Text Available As many nuclear power plants are reaching their end of lifecycle, the decommissioning of these installations has become one of the 21st century’s great challenges. Each project may be managed differently, depending on the country, development policies, financial considerations, and the availability of qualified engineers or specialized companies to handle such projects. The principle objective of decommissioning is to place a facility into such a condition that there is no unacceptable risk from the decommissioned facility to public health and safety of the environment. In order to ensure that at the end of its life the risk from a facility is within acceptable bounds, action is normally required. The overall decommissioning strategy is to deliver a timely, cost-effective program while maintaining high standards of safety, security and environmental protection. If facilities were not decommissioned, they could degrade and potentially present an environmental radiological hazard in the future. Simply abandoning or leaving a facility after ceasing operations is not considered to be an acceptable alternative to decommissioning. The final aim of decommissioning is to recover the geographic site to its original condition.

  13. Operating Experience of MACSTOR Modules at CANDU 6 Stations

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, Robert R. [Atomic Energy Canada Ltd., Chalk River (Canada)

    2005-11-15

    Over the last three decades, Atomic Energy of Canada Limited (AECL) has contributed to the technology development and implementation of dry spent fuel management facilities in Canada, Korea and Romania During that period, AECL has developed a number of concrete canister models and the MACSTOR200 module, a medium size air-cooled vault with a 228 MgU (Mega grams of Uranium) capacity. AECL's dry storage technologies were used for the construction of eight large-scale above ground dry storage facilities for CANDU spent fuel. As of 2005, those facilities have an installed capacity in excess of 5,000 MgU. Since 1995, the two newest dry storage installations built for CANDU 6 reactors at Gentilly 2 (Canada) and Cernavoda (Romania) used the MACSTOR 200 module. Seven such modules have been built at Gentilly 2 during the 1995 to 2004 period and one at Cernavoda in 2003. The construction and operating experience of those modules is reviewed in this paper. The MACSTOR 200 modules were initially designed for a 50-year service life, with recent units at Gentilly 2 licensed for a 100-year service life in a rural (non-maritime) climate. During the 1995-2005 period, six of the eight modules were loaded with fuel. Their operation has brought a significant amount of experience on loading operations, performance of fuel handling equipment, radiation shielding, heat transfer, monitoring of the two confinement boundaries and radiation dose to personnel. Heat dissipation performance of the MACSTOR 200 was initially licensed using values derived from full scale tests made at AECL's Whiteshell Research Laboratories, that were backed-up by temperature measurements made on the first two modules. Results and computer models developed for the MACSTOR 200 module are described. Korea Hydro and Nuclear Power (KHNP) and its subsidiary Nuclear Environment Technology Institute (NETEC), in collaboration with Hyundai Engineering Company Ltd. (HEC) and AECL, are developing a new dry storage

  14. A review of CANDU feeder wall thinning

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Han Sub [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Flow Accelerated Corrosion is an active degradation mechanism of CANDU feeder. The tight bend downstream to Grayloc weld connection, close to reactor face, suffers significant wall thinning by FAC. Extensive in-service inspection of feeder wall thinning is very difficult because of the intense radiation field, complex geometry, and space restrictions. Development of a knowledge-based inspection program is important in order to guarantee that adequate wall thickness is maintained throughout the whole life of feeder. Research results and plant experiences are reviewed, and the plant inspection databases from Wolsong Units One to Four are analyzed in order to support developing such a knowledge-based inspection program. The initial thickness before wall thinning is highly non-uniform because of bending during manufacturing stage, and the thinning rate is non-uniform because of the mass transfer coefficient distributed non-uniformly depending on local hydraulics. It is obvious that the knowledge-based feeder inspection program should focus on both fastest thinning locations and thinnest locations. The feeder wall thinning rate is found to be correlated proportionately with QV of each channel. A statistical model is proposed to assess the remaining life of each feeder using the QV correlation and the measured thicknesses. W-1 feeder suffered significant thinning so that the shortest remaining life barely exceeded one year at the end of operation before replacement. W-2 feeder showed far slower thinning than W-1 feeder despite the faster coolant flow. It is believed that slower thinning in W-2 is because of higher chromium content in the carbon steel feeder material. The average Cr content of W-2 feeder is 0.051%, while that value is 0.02% for W-1 feeder. It is to be noted that FAC is reduced substantially even though the Cr content of W-2 feeder is still very low

  15. Development of CANDU pressure tube integrity evaluation system

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  17. Future CANDU nuclear power plant design requirements document executive summary

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of); Usmani, S.A. [Atomic Energy of Canada Ltd., Toronto (Canada)

    1996-03-01

    The future CANDU Requirements Document (FCRED) describes a clear and complete statement of utility requirements for the next generation of CANDU nuclear power plants including those in Korea. The requirements are based on proven technology of PHWR experience and are intended to be consistent with those specified in the current international requirement documents. Furthermore, these integrated set of design requirements, incorporate utility input to the extent currently available and assure a simple, robust and more forgiving design that enhances the performance and safety. The FCRED addresses the entire plant, including the nuclear steam supply system and the balance of the plant, up to the interface with the utility grid at the distribution side of the circuit breakers which connect the switchyard to the transmission lines. Requirements for processing of low level radioactive waste at the plant site and spent fuel storage requirements are included in the FCRED. Off-site waste disposal is beyond the scope of the FCRED. 2 tabs., 1 fig. (Author) .new.

  18. Optimization and implementation study of plutonium disposition using existing CANDU Reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    Since early 1994, the Department of Energy has been sponsoring studies aimed at evaluating the merits of disposing of surplus US weapons plutonium as Mixed Oxide (MOX) fuel in existing commercial Canadian Pressurized Heavy Water reactors, known as CANDU`s. The first report, submitted to DOE in July, 1994 (the 1994 Executive Summary is attached), identified practical and safe options for the consumption of 50 to 100 tons of plutonium in 25 years in some of the existing CANDU reactors operating the Bruce A generating station, on Lake Huron, about 300 km north east of Detroit. By designing the fuel and nuclear performance to operate within existing experience and operating/performance envelope, and by utilizing existing fuel fabrication and transportation facilities and methods, a low cost, low risk method for long term plutonium disposition was developed. In December, 1995, in response to evolving Mission Requirements, the DOE requested a further study of the CANDU option with emphasis on more rapid disposition of the plutonium, and retaining the early start and low risk features of the earlier work. This report is the result of that additional work.

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Yong [Korea Electric Power Corporation Research Institute, Daejeon (Korea, Republic of); Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of)

    2011-08-15

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

  20. Pre-licensing of the Advanced CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ion, R.; Popov, N.K.; Snell, V. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)]. E-mail: popovn@aecl.ca; West, J. [Candesco, Toronto, Ontario (Canada); Xu, C. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2006-07-01

    Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving 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. AECL initiated pre-licensing reviews of the ACR reactor design in Canada, US and China, with an objective to take into account regulatory feedback early in the design process. The Canadian Nuclear Safety Commission (CNSC) is performing a pre-project pre-licensing assessment of the ACR design. The objective of the assessment is to issue a formal statement as to whether there are any fundamental barriers that would prevent the licensing of the new CANDU reactor design in Canada under the Nuclear Safety and Control Act. The CNSC review is being conducted in four phases. In Phase 1 (September 2003 to September 2004) CNSC performed a pre-licensing review of the ACR-700, and focused on the design process, methodology, design concepts and R and D. CNSC staff reviewed about 100 reports, and submitted to AECL questions and comments. In Phase 2 (September 2004 to August 2005) AECL provided responses and additional information to CNSC on their comments and questions in Phase 1. Phase 3 is the Transition Phase (September 2005 to May 2006), bridging the transition from the ACR-700 to the ACR-1000 design. Phase 3 focused on review of generic aspects of the ACR design, on the Safety

  1. The Effects of the PEM Fuel Cell Performance with the Waved Flow Channels

    OpenAIRE

    Yue-Tzu Yang; Kuo-Teng Tsai; Cha’o-Kuang Chen

    2013-01-01

    The objective of this study is to use a new style of waved flow channel instead of the plane surface channel in the proton exchange membrane fuel cell (PEMFC). The velocity, concentration, and electrical performance with the waved flow channel in PEMFC are investigated by numerical simulations. The results show that the waved channel arises when the transport benefits through the porous layer and improves the performance of the PEMFC. This is because the waved flow channel enhances the forced...

  2. Once-through CANDU reactor models for the ORIGEN2 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A.G.; Bjerke, M.A.

    1980-11-01

    Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % /sup 235/U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given.

  3. Failure probability estimation of flaw in CANDU pressure tube considering the dimensional change

    Energy Technology Data Exchange (ETDEWEB)

    Kwak, Sang Log; Kim, Young Jin [Sungkyunkwan Univ., Suwon (Korea, Republic of); Lee, Joon Seong [Kyonggi Univ., Suwon (Korea, Republic of); Park, Youn Won [KINS, Taejon (Korea, Republic of)

    2002-11-01

    The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and heavy water coolant. Pressure tubes are installed horizontally inside the reactor and only selected samples are periodically examined during in-service inspection. In this respect, a probabilistic safety assessment method is more appropriate for the assessment of overall pressure tube safety. The failure behavior of CANDU pressure tubes, however, is governed by delayed hydride cracking which is the major difference from pipings and reactor pressure vessels. Since the delayed hydride cracking has more widely distributed governing parameters, it is impossible to apply a general PFM methodology directly. In this paper, a PFM methodology for the safety assessment of CANDU pressure tubes is introduced by applying Monte Carlo simulation in determining failure probability. Initial hydrogen concentration, flaw shape and depth, axial and radial crack growth rate and fracture toughness were considered as probabilistic variables. Parametric study has been done under the base of pressure tube dimension and hydride precipitation temperature in calculating failure probability. Unstable fracture and plastic collapse are used for the failure assessment. The estimated failure probability showed about three-order difference with changing dimensions of pressure tube.

  4. A generalized perturbation program for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Yang, Won Sik [Chosun University, Kwangju (Korea, Republic of)

    1998-12-31

    A generalized perturbation program has been developed for the purpose of estimating zonal power variation of a CANDU reactor upon refueling operation. The forward and adjoint calculation modules of RFSP code were used to construct the generalized perturbation program. The numerical algorithm for the generalized adjoint flux calculation was verified by comparing the zone power estimates upon refueling with those of forward calculation. It was, however, noticed that the truncation error from the iteration process of the generalized adjoint flux is not negligible. 2 refs., 1 figs., 1 tab. (Author)

  5. Radiological Characteristics of decommissioning waste from a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-11-15

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

  6. Detection of Liquid Water in PEM Fuel Cells' Channels: Design and Validation of a Microsensor.

    OpenAIRE

    Conteau, Delphine; Bonnet, Caroline; Funfschilling, Denis; Weber, Mathieu; Didierjean, Sophie; Lapicque, François

    2010-01-01

    Abstract Suitable water management is a critical issue to reach the full potential of PEM fuel cells: whereas the membrane must be hydrated enough, liquid droplets formed by water in excess can block the flow in the gas distribution channels and hinder the fuel cell performance. In order to detect the presence of droplets in cathode flow channel, an electrochemical sensor has been developed and tested in a dedicated emulation cell. It is based on the widely used principle of two-el...

  7. Development of bipolar plates with different flow channel configurations for fuel cells

    Science.gov (United States)

    Boddu, Rajesh; Marupakula, Uday Kumar; Summers, Benjamin; Majumdar, Pradip

    Bipolar plates include separate gas flow channels for anode and cathode electrodes of a fuel cell. These gases flow channels supply reactant gasses as well as remove products from the cathode side of the fuel cell. Fluid flow, heat and mass transport processes in these channels have significant effect on fuel cell performance, particularly to the mass transport losses. The design of the bipolar plates should minimize plate thickness for low volume and mass. Additionally, contact faces should provide a high degree of surface uniformity for low thermal and electrical contact resistances. Finally, the flow fields should provide for efficient heat and mass transport processes with reduced pressure drops. In this study, bipolar plates with different serpentine flow channel configurations are analyzed using computational fluid dynamics modeling. Flow characteristics including variation of pressure in the flow channel across the bipolar plate are presented. Pressure drop characteristics for different flow channel designs are compared. Results show that with increased number of parallel channels and smaller sizes, a more effective contact surface area along with decreased pressured drop can be achieved. Correlations of such entrance region coefficients will be useful for the PEM fuel cell simulation model to evaluate the affects of the bipolar plate design on mass transfer loss and hence on the total current and power density of the fuel cell.

  8. Pressure tube creep impact on the physics parameters for CANDU-6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, W. Y.; Min, B. J. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kam, S. C.; Kim, M. E. [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2004-07-01

    The lattice cell calculations are performed to assess the sensitivity of the reactor physics parameters to pressure tube creep resulting from radiation aging. The physics parameters of the lattice cell are calculated by using WIMSD-5B code, WIMS- AECL code, and MCNP code. The reference model(normal state) and two perturbed models accounting for the pressure tube creep are developed on the basis of CANDU-6 lattice cell. The 2.5% and 5% values of pressure tube diameter creep are considered. Also, The effects of the analyzed lattice parameters which are the coolant void reactivity, the fuel fission density and the atom density of Pu isotopes on the lattice.

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

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

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

  10. EC6{sup TM} - Enhanced Candu 6{sup TM} reactor safety characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.; Snell, V.; Cormier, M.; Hopwood, J. [Atomic Energy of Canada Ltd., 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (Canada)

    2010-07-01

    The EC6 is a 740 MWe-class natural-uranium-fuelled, heavy-water-cooled and -moderated pressure-tube reactor, which has evolved from the eleven (11) CANDU{sup R} 6 plants operating in five countries (on four continents). CANDU 6 has over 150 reactor-years of safe operation. The most recent CANDU 6 - at Qinshan, in China - is the Reference Design for EC6. The EC6 shares many inherent, passive and engineered safety characteristics with the Reference Design. However EC6 has been designed to meet modern regulatory requirements and safety expectations. The resulting design changes have improved these safety characteristics, and this paper provides a convenient summary. The paper addresses the safety functions of reactivity control, heat removal, and containment of radioactive material. For each safety function, the EC6 characteristics are categorized as inherent, passive, or engineered. The paper focuses mostly on the first two. The Enhanced CANDU 6 uses an appropriate mix of passive, inherent, and engineered safety functions. Reactivity transients are generally slow, mild and inherently limited due to the natural uranium core and use of on-power refuelling. Only the coolant void coefficient can cause a large reactivity insertion, particularly in a large LOCA. This is mitigated by the long prompt neutron lifetime and the large delayed neutron fraction, and terminated by either of the two shutdown systems. For EC6, the large LOCA power transient has been reduced significantly by speeding up the slower of the two shutdown systems. Redundant shutdown and the LOCA power pulse improvements mitigate the limiting large positive reactivity insertion. Decay heat removal shows a very high component of passive safety, from thermo-siphoning in the Reactor Coolant System to passive heat removal in severe accidents via the moderator or reactor vault. The latter two can maintain the fuel in a more predictable and favourable geometry than 'core on the floor'. The containment

  11. Speciation of iodine (I-127) in the natural environment around Canadian CANDU sites

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, S.J.; Kotzer, T.G.; Chant, L.A

    2001-06-01

    In Canada, very little data is available regarding the concentrations and chemical speciation of iodine in the environment proximal and distal to CANDU Nuclear Power Generating Stations (NPGS). In the immediate vicinity of CANDU reactors, the short-lived iodine isotope {sup 131}I (t{sub 1/2} = 8.04 d), which is produced from fission reactions, is generally below detection and yields little information about the environmental cycling of iodine. Conversely, the fission product {sup 129}I has a long half-life (t{sub 1/2} = 1.57x10{sup 7} y) and has had other anthropogenic inputs (weapons testing, nuclear fuel reprocessing) other than CANDU over the past 50 years. As a result, the concentrations of stable iodine ({sup 127}I) have been used as a proxy. In this study, a sampling system was developed and tested at AECL's Chalk River Laboratories (CRL) to collect and measure the particulate and gaseous inorganic and organic fractions of stable iodine ({sup 127}I) in air and associated organic and inorganic reservoirs. Air, vegetation and soil samples were collected at CRL, and at Canadian CANDU Nuclear Power Generating Stations (NPGS) at OPG's (Ontario Power Generation) Pickering (PNGS) and Darlington NPGS (DNGS) in Ontario, as well as at NB Power's Pt. Lepreau NPGS in New Brunswick. The concentrations of particulate and inorganic iodine in air at CRL were extremely low, and were often found to be below detection. The concentrations are believed to be at this level because the sediments in the CRL area are glacial fluvial and devoid of marine ionic species, and the local atmospheric conditions at the sampling site are very humid. Concentrations of a gaseous organic species were comparable to worldwide levels. The concentrations of particulate and inorganic iodine in air were also found to be low at PNGS and DNGS, which may be attributed to reservoir effects of the large freshwater lakes in southern Ontario, which might serve to dilute the atmospheric iodine

  12. Investigation of low flow trip alarm of FARE channel at Wolsong NPP

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S. K.; Lee, S. G.; Son, S. M.; Lee, W. R.; Kim, T. R. [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Lee, K. Y.; Hong, S. S. [Korea Hydro and Nuclear Power Co., Ltd., Gyeongju (Korea, Republic of); Namgung, I. [KOPEC, Taejon (Korea, Republic of)

    2003-10-01

    At Wolsong site in Korea, four CANDU6 type reactors are operating now. The CANDU type reactors require on power refuelling by the two remotely controlled F/Ms (Fuelling Machines). The refuelling operation is carried out along the coolant direction by the drag force of fuel bundles for most channels; however, for about 30% of channels the coolant flow is not sufficiently high enough to carry fuel bundles to the downstream. Therefore, a device, FARE(Flow Assist Ram Extension) device, is used to create an additional drag force to push the fuel bundles along the coolant flow. The FARE device has been causing the channel flow low phenomena of SDS1 (shut down system 1) channels due to more than excessive blockage of coolant flow. The flow rate change for the SDS1 channels is monitored on line for the coolant flow blockage of the channels. This has initiated the investigation of the FARE device behavior and the cause and remedy for the problem. The analysis revealed that the channel flow low occurs as the insertion of the FARE device into the active core flow and disappears as the removal of the FARE device from it. In order to eliminate the channel flow low phenomena, flow resistance by the FARE device should be reduced. The analysis has revealed that a too much relaxation of flow resistance would make the FARE device unable to push the fuel bundle string, and a too little relaxation would make the FARE device unable to eliminate the channel flow low phenomena. Suggested design changes include more flow holes in the tube.

  13. Development of CANFLEX fuel fabrication technology

    Energy Technology Data Exchange (ETDEWEB)

    Kang, M. S.; Choi, C. B.; Park, C. H.; Kwon, W. J.; Kim, C. H.; Kim, B. J.; Koo, C. H.; Cho, D. S.; So, D. Y.; Suh, S. W.; Park, C. J.; Chang, D. H.; Yun, S. H. [KEPCO Nuclear Fuel Company, Taejeon (Korea)

    2000-04-01

    Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU(CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. One of the improvements in CANDU fuel fabrication technology, and advanced method of Zr-Be brazing was developed. For the formation of Zr-Be alloy, preheating and main heating temperature in the furnace is 700 deg C, 1200 deg C respectively. In order to find an appropriate material for the brazing joints in the CANDU fuel, the composition of Zr based amorphous metals were designed. And, the effect of hydrogen on the mechanical properties of cladding sheath and feasibility of the eddy current test to evaluate quality of end cap weld were also studied for the fundamental research purpose. As a preliminary study to suggest optimal way for the mass production of CANFLEX-NU fuel at KNFC the existing CANDU fuel facilities and fabrication/inspection processes were reviewed. The best way is that the current CANDU facility shall be modified to produce small diametrial CANFLEX elements and a new facility shall be constructed to produce large diametrial CANFLEX fuel elements. 46 refs., 99 figs., 10 tabs. (Author)

  14. Development of an Integrity Assessment Procedure for CANDU Pressure Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Han Sub [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    The pressure tubes used in a CANDU reactor are made from Zr-2.5Nb. During service the pressure tubes operate at temperatures between about 150 and 310 .deg. C, and with variable coolant pressures up to 11MPa corresponding to hoop stress of up to 130MPa. The maximum flux of fast neutrons (E>1MeV) from the fuel is about 4X10{sup 17}nm{sup -2}{sub s}{sup -1}. The pressure tubes are exposed to very severe degradation environment. The aging degradation of the pressure tubes are summarized as below. - Geometric deformation; axial elongation, diametric creep, and wall thinning. - Deuterium uptake; some fraction of the deuterium generated by the corrosion of pressure tubes is absorbed into the pressure tubes. Total equivalent hydrogen content in the pressure tube is the sum of the initial hydrogen content before operation and the deuterium uptake during operation. High concentration of hydrogen inside the pressure tubes makes the metal susceptible to Delayed Hydride Cracking. The DHC is a degradation mechanism of prime importance for CANDU pressure tubes. Mechanical properties, in particular fracture toughness, are deteriorated by high concentration of dissolved hydrogen. - Flaws; volumetric flaws are generated during operation. Wear scars by debris fretting, and bearing pad fretting are common. These volumetric flaws can be a site of crack initiation by fatigue or DHC. Cracks can propagate by DHC or fatigue crack propagation if conditions are met. - Material properties degradation; mechanical properties are affected by neutron irradiation. Yield strength and tensile strength are increased, and fracture toughness is deteriorated. The susceptibility to DHC is also affected. The integrity assessment of the pressure tube is a procedure to determine if the risk of pressure tube failure is controlled to maintain acceptably low. CSA N285.4 and 285.8 are two important guidelines regarding the integrity of pressure tubes. N285.4 is to guide in-service inspection, and N285

  15. Advanced CANDU reactor pre-licensing progress

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.K.; West, J.; Snell, V.G.; Ion, R.; Archinoff, G.; Xu, C. [Atomic Energy of Canada Limited., Mississauga, Ontario (Canada)]. E-mail: popovn@aecl.ca

    2005-07-01

    The Advanced CANDU Reactor (ACR) is an evolutionary advancement of the current CANDU 6 reactor, aimed at producing electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The Canadian Nuclear Safety Commission (CNSC) staff are currently reviewing the ACR design to determine whether, in their opinion, there are any fundamental barriers that would prevent the licensing of the design in Canada. This CNSC licensability review will not constitute a licence, but is expected to reduce regulatory risk. The CNSC pre-licensing review started in September 2003, and was focused on identifying topics and issues for ACR-700 that will require a more detailed review. CNSC staff reviewed about 120 reports, and issued to AECL 65 packages of questions and comments. Currently CNSC staff is reviewing AECL responses to all packages of comments. AECL has recently refocused the design efforts to the ACR-1000, which is a larger version of the ACR design. During the remainder of the pre-licensing review, the CNSC review will be focused on the ACR-1000. AECL Technologies Inc. (AECLT), a wholly-owned US subsidiary of AECL, is engaged in a pre-application process for the ACR-700 with the US Nuclear Regulatory Commission (USNRC) to identify and resolve major issues prior to entering a formal process to obtain standard design certification. To date, the USNRC has produced a Pre-Application Safety Assessment Report (PASAR), which contains their reviews of key focus topics. During the remainder of the pre-application phase, AECLT will address the issues identified in the PASAR. Pursuant to the bilateral agreement between AECL and the Chinese nuclear regulator, the National Nuclear Safety Administration (NNSA) and its Nuclear Safety Center (NSC), NNSA/NSC are reviewing the ACR in seven focus areas. The review started in September 2004, and will take three years. The main objective of the review is to determine how the ACR complies

  16. Leak detection capability in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  17. Water behavior in a u-shaped flow channel of PEM fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Quan, P.; Zhou, B.; Sobiesiak, A. [Windsor Univ., ON (Canada). Dept. of Mechanical, Automotive and Materials Engineering; Liu, Z.S. [National Research Council of Canada, Vancouver, BC (Canada). Inst. for Fuel Cell Innovation

    2005-07-01

    A study was conducted to find a practical approach for predicting liquid water distribution in the U-shaped flow channels of a proton exchange membrane (PEM) fuel cell. Computational fluid dynamics modeling with the FLUENT software package was used to demonstrate the two-phase flow of the air-water transport process inside the channel. It was noted that no chemical reaction occurs inside the flow channels and the liquid water is formed either on the surfaces of the flow channels or inside the flow channels. The problem can therefore be simplified as a fluid mechanics problem with water sources inside its physical domain or on its boundaries. The volume-of-fluid (VOF) model was used to track dynamic air-water interactions. Three cases with a range of initial water phase distributions corresponding to different fuel cell operating conditions were simulated numerically to gain a better understanding of water behaviour inside the serpentine channel. It was concluded that the bend area in the serpentine flow field affects the fuel cell performance. This is because it influences the flow field which in turn influences the air-water flow and water liquid distribution inside the channel or along the inside channel surfaces. 15 refs., 1 tab., 11 figs.

  18. Two phase flow simulation in a channel of a polymer electrolyte membrane fuel cell using the lattice Boltzmann method

    OpenAIRE

    Ben Salah, Yasser; Tabe, Yutaka; Chikahisa, Takemi

    2012-01-01

    Water management in polymer electrolyte (PEM) fuel cells is important for fuel cell performance and durability. Numerical simulations using the lattice Boltzmann method (LBM) are developed to elucidate the dynamic behavior of condensed water and gas flows in a polymer electrolyte membrane (PEM) fuel cell gas channel. A scheme for two-phase flow with large density differences was applied to establish the optimum gas channel design for different gas channel heights, droplet positions, and gas c...

  19. DUPIC fuel compatibility assessment

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Rho, G. H.; Park, J. W. [and others

    2000-03-01

    The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition.

  20. Chemistry aspects of the source term formation for a severe accident in a CANDU type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Constantin, A.; Constantin, M. [Institute for Nuclear Research, Pitesti (Romania)

    2013-07-15

    The progression of a severe accident in a CANDU type reactor is slow because the core is surrounded by a large quantity of heavy and light water which acts as a heat sink to remove the decay heat. Therefore, the source term formation is a complex and long process involving fission products transport and releasing in the fuel matrix, thermal hydraulics of the transport fluid in the primary heat system and containment, deposition and transport of fission products, chemistry including the interaction with the dousing system, structural materials and paints, etc. The source term is strongly dependent on initial conditions and accident type. The paper presents chemistry aspects for a severe accident in a CANDU type reactor, in terms of the retention in the primary heat system. After releasing from the fuel elements, the fission products suffer a multitude of phenomena before they are partly transferred into the containment region. The most important species involved in the deposition were identified. At the same time, the influence of the break position in the transfer fractions from the primary heat system to the containment was investigated. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-10-01

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

  2. A flow channel design procedure for PEM fuel cells with effective water removal

    Science.gov (United States)

    Li, Xianguo; Sabir, Imran; Park, Jaewan

    Proper water management in polymer electrolyte membrane (PEM) fuel cells is critical to achieve the potential of PEM fuel cells. Membrane electrolyte requires full hydration in order to function as proton conductor, often achieved by fully humidifying the anode and cathode reactant gas streams. On the other hand, water is also produced in the cell due to electrochemical reaction. The combined effect is that liquid water forms in the cell structure and water flooding deteriorates the cell performance significantly. In the present study, a design procedure has been developed for flow channels on bipolar plates that can effectively remove water from the PEM fuel cells. The main design philosophy is based on the determination of an appropriate pressure drop along the flow channel so that all the liquid water in the cell is evaporated and removed from, or carried out of, the cell by the gas stream in the flow channel. At the same time, the gas stream in the flow channel is maintained fully saturated in order to prevent membrane electrolyte dehydration. Sample flow channels have been designed, manufactured and tested for five different cell sizes of 50, 100, 200, 300 and 441 cm 2. Similar cell performance has been measured for these five significantly different cell sizes, indicating that scaling of the PEM fuel cells is possible if liquid water flooding or membrane dehydration can be avoided during the cell operation. It is observed that no liquid water flows out of the cell at the anode and cathode channel exits for the present designed cells during the performance tests, and virtually no liquid water content in the cell structure has been measured by the neutron imaging technique. These measurements indicate that the present design procedure can provide flow channels that can effectively remove water in the PEM fuel cell structure.

  3. A flow channel design procedure for PEM fuel cells with effective water removal

    Energy Technology Data Exchange (ETDEWEB)

    Li, Xianguo; Sabir, Imran; Park, Jaewan [Department of Mechanical Engineering, University of Waterloo, Waterloo, Ontario N2L 3G1 (Canada)

    2007-01-01

    Proper water management in polymer electrolyte membrane (PEM) fuel cells is critical to achieve the potential of PEM fuel cells. Membrane electrolyte requires full hydration in order to function as proton conductor, often achieved by fully humidifying the anode and cathode reactant gas streams. On the other hand, water is also produced in the cell due to electrochemical reaction. The combined effect is that liquid water forms in the cell structure and water flooding deteriorates the cell performance significantly. In the present study, a design procedure has been developed for flow channels on bipolar plates that can effectively remove water from the PEM fuel cells. The main design philosophy is based on the determination of an appropriate pressure drop along the flow channel so that all the liquid water in the cell is evaporated and removed from, or carried out of, the cell by the gas stream in the flow channel. At the same time, the gas stream in the flow channel is maintained fully saturated in order to prevent membrane electrolyte dehydration. Sample flow channels have been designed, manufactured and tested for five different cell sizes of 50, 100, 200, 300 and 441cm{sup 2}. Similar cell performance has been measured for these five significantly different cell sizes, indicating that scaling of the PEM fuel cells is possible if liquid water flooding or membrane dehydration can be avoided during the cell operation. It is observed that no liquid water flows out of the cell at the anode and cathode channel exits for the present designed cells during the performance tests, and virtually no liquid water content in the cell structure has been measured by the neutron imaging technique. These measurements indicate that the present design procedure can provide flow channels that can effectively remove water in the PEM fuel cell structure. (author)

  4. THE IMPACT OF POWER COEFFICIENT OF REACTIVITY ON CANDU 6 REACTORS

    Directory of Open Access Journals (Sweden)

    D. KASTANYA

    2013-10-01

    Full Text Available The combined effects of reactivity coefficients, along with other core nuclear characteristics, determine reactor core behavior in normal operation and accident conditions. The Power Coefficient of Reactivity (PCR is an aggregate indicator representing the change in reactor core reactivity per unit change in reactor power. It is an integral quantity which captures the contributions of the fuel temperature, coolant void, and coolant temperature reactivity feedbacks. All nuclear reactor designs provide a balance between their inherent nuclear characteristics and the engineered reactivity control features, to ensure that changes in reactivity under all operating conditions are maintained within a safe range. The CANDU® reactor design takes advantage of its inherent nuclear characteristics, namely a small magnitude of reactivity coefficients, minimal excess reactivity, and very long prompt neutron lifetime, to mitigate the demand on the engineered systems for controlling reactivity and responding to accidents. In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with their design characteristics, such that the overall design and safety characteristics of the reactor are not sensitive to the value of the PCR. For other reactor design concepts a PCR which is both large and negative is an important aspect in the design of their engineered systems for controlling reactivity. It will be demonstrated that during Loss of Regulation Control (LORC and Large Break Loss of Coolant Accident (LBLOCA events, the impact of variations in power coefficient, including a hypothesized larger than estimated PCR, has no safety-significance for CANDU reactor design. Since the CANDU 6 PCR is small, variations in the range of values for PCR on the performance or safety of the reactor are not significant.

  5. Optimization of polymer electrolyte membrane fuel cell flow channels using a genetic algorithm

    Science.gov (United States)

    Catlin, Glenn; Advani, Suresh G.; Prasad, Ajay K.

    The design of the flow channels in PEM fuel cells directly impacts the transport of reactant gases to the electrodes and affects cell performance. This paper presents results from a study to optimize the geometry of the flow channels in a PEM fuel cell. The optimization process implements a genetic algorithm to rapidly converge on the channel geometry that provides the highest net power output from the cell. In addition, this work implements a method for the automatic generation of parameterized channel domains that are evaluated for performance using a commercial computational fluid dynamics package from ANSYS. The software package includes GAMBIT as the solid modeling and meshing software, the solver FLUENT, and a PEMFC Add-on Module capable of modeling the relevant physical and electrochemical mechanisms that describe PEM fuel cell operation. The result of the optimization process is a set of optimal channel geometry values for the single-serpentine channel configuration. The performance of the optimal geometry is contrasted with a sub-optimal one by comparing contour plots of current density, oxygen and hydrogen concentration. In addition, the role of convective bypass in bringing fresh reactant to the catalyst layer is examined in detail. The convergence to the optimal geometry is confirmed by a bracketing study which compares the performance of the best individual to those of its neighbors with adjacent parameter values.

  6. Beyond designed functional margins in CANDU type NPP. Radioactive nuclei assessment in an LOCA type accident

    Directory of Open Access Journals (Sweden)

    Budu Andrei Razvan

    2015-01-01

    Full Text Available European Union's energy roadmap up to year 2050 states that in order to have an efficient and sustainable economy, with minimum or decreasing greenhouse gas emissions, along with use of renewable resources, each constituent state has the option for nuclear energy production as one desirable option. Every scenario considered for tackling climate change issues, along with security of supply positions the nuclear energy as a recommended option, an option that is highly competitive with respect to others. Nuclear energy, along with other renewable power sources are considered to be the main pillars in the energy sector for greenhouse gas emission mitigation at European level. European Union considers that nuclear energy must be treated as a highly recommended option since it can contribute to security of energy supply. Romania showed excellent track-records in operating in a safe and economically sound manner of Cernavoda NPP Units 1&2. Both Units are in top 10 worldwide in terms of capacity factor. Due to Romania's need to ensure the security of electricity supply, to meet the environmental targets and to move to low carbon generation technologies, Cernavoda Units 3&4 Project appears as a must. This Project was started in 2010 and it is expected to have the Units running by 2025. Cost effective and safety operation of a Nuclear Power Plant is made taking into consideration functional limits of its equipment. As common practice, every nuclear reactor type (technology used is tested according to the worse credible accident or equipment failure that can occur. For CANDU type reactor, this is a Loss of Cooling Accident (LOCA. In a LOCA type accident in a CANDU NPP, using RELAP/SCDAP code for fuel bundle damage assessment the radioactive nuclei are to be quantified. Recently, CANDU type NPP accidents are studied using the RELAP/SCDAP code only. The code formerly developed for PWR type reactors was adapted for the CANDU geometry and can assess the

  7. Radionuclide Release after End Fitting Failure Accident in CANDU-6 Plant

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hoon; Kim, Yun Ho; Lee, Kwang Ho [Korea Electric Power Corporation Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    The total amount of mass and energy discharged into containment building from primary heat transport system in the event of end fitting failure accident in CANDU-6 plant is similar to that of small loss of coolant accident. But the ejection of fuel bundles into fuelling machine room is unique phenomenon and causes radio nuclides release from the physically broken fuel rod to outside containment building. The only objective of containment behavior analysis for end fitting failure event is to assess the amount of radio nuclides release to the ambient atmosphere. Radionuclide release rates in case of end fitting failure with all safety system available, that is containment building is intact, as well as with containment system impairment are analyzed with GOTHIC and SMART code

  8. Direct reuse of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Nader M.A., E-mail: mnader73@yahoo.com

    2014-10-15

    Highlights: • A new design for the PWR assemblies for direct use of spent fuel was proposed. • The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors. • The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. • MCNPX is used for the calculations that showed that the burnup can be increased by about 25%. • Acceptable linear heat generation rate in hot rods and improved Pu proliferation resistance. - Abstract: In this paper we proposed a new design for the PWR fuel assembly for direct use of the PWR spent fuel without processing. The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors which preferably built in the same site to avoid the problem of transportations. The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. Each tube has the same inner diameter of that of CANDU pressure tube. The spaces between the tubes contain low enriched UO{sub 2} fuel rods and guide tubes. MCNPX code is used for the simulation and calculation of the burnup of the proposed assembly. The bundles after the discharge from the PWR with their materials inventories are burned in a CANDU cell after a certain decay time. The results were compared with reference results and the impact of this new design on the uranium utilization improvement and on the proliferation resistance of plutonium is discussed. The effect of this new design on the power peaking, moderator temperature coefficient of reactivity and CANDU coolant void reactivity are discussed as well.

  9. Incorporating single detector failure into the ROP detector layout optimization for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kastanya, Doddy, E-mail: Doddy.Kastanya@snclavalin.com

    2015-12-15

    Highlights: • ROP TSP value needs to be adjusted when any detector in the system fails. • Single detector failure criterion has been incorporated into the detector layout optimization as a constraint. • Results show that the optimized detector layout is more robust with respect to its vulnerability to a single detector failure. • An early rejection scheme has been introduced to speed-up the optimization process. - Abstract: In CANDU{sup ®} reactors, the regional overpower protection (ROP) systems are designed to protect the reactor against overpower in the fuel which could reduce the safety margin-to-dryout. In the CANDU{sup ®} 600 MW (CANDU 6) design, there are two ROP systems in the core, each of which is connected to a fast-acting shutdown system. Each ROP system consists of a number of fast-responding, self-powered flux detectors suitably distributed throughout the core within vertical and horizontal flux detector assemblies. The placement of these ROP detectors is a challenging discrete optimization problem. In the past few years, two algorithms, DETPLASA and ADORE, have been developed to optimize the detector layout for the ROP systems in CANDU reactors. These algorithms utilize the simulated annealing (SA) technique to optimize the placement of the detectors in the core. The objective of the optimization process is typically either to maximize the TSP value for a given number of detectors in the system or to minimize the number of detectors in the system to obtain a target TSP value. One measure to determine the robustness of the optimized detector layout is to evaluate the maximum decrease (penalty) in TSP value when any single detector in the system fails. The smaller the penalty, the more robust the design is. Therefore, in order to ensure that the optimized detector layout is robust, the single detector failure (SDF) criterion has been incorporated as an additional constraint into the ADORE algorithm. Results from this study indicate that there

  10. Numerical studies on liquid water flooding in gas channels used inpolymer electrolyte fuel cells

    NARCIS (Netherlands)

    Qin, CZ.; Hassanizadeh, S.M.; Rensink, D.

    2012-01-01

    Water management plays an important role in the development of low-temperature polymer electrolyte fuel cells (PEFCs). The lack of a macroscopic gas channel (GC) flooding model constrains the current predictions of PEFC modeling under severe flooding situations. In this work, we have extended our pr

  11. Mathematical modeling of channel-porous layer interfaces in PEM fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Ehrhardt, M.; Fuhrmann, J.; Holzbecher, E.; Linke, A.

    2008-07-01

    In proton exchange membrane (PEM) fuel cells, the transport of the fuel to the active zones, and the removal of the reaction products are realized using a combination of channels and porous diffusion layers. In order to improve existing mathematical and numerical models of PEM fuel cells, a deeper understanding of the coupling of the flow processes in the channels and diffusion layers is necessary. After discussing different mathematical models for PEM fuel cells, the work focuses on the description of the coupling of the free flow in the channel region with the filtration velocity in the porous diffusion layer as well as interface conditions between them. The difficulty in finding effective coupling conditions at the interface between the channel flow and the membrane lies in the fact that often the orders of the corresponding differential operators are different, e.g., when using stationary (Navier-)Stokes and Darcy's equation. Alternatively, using the Brinkman model for the porous media this difficulty does not occur. We review different interface conditions, including the well-known Beavers-Joseph-Saffman boundary condition and its recent improvement by Le Bars and Worster. (orig.)

  12. CANDU steam generator life management: laboratory data and plant experience

    Energy Technology Data Exchange (ETDEWEB)

    Tapping, R.L. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Nickerson, J.H.; Subash, N. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Wright, M.D

    2001-10-01

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

  13. A study for good regulatin of the CANDU's in Korea. Development of safety regulatory requirement for CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se Ki; Shin, Y. K.; Kim, J. S.; Yu, Y. J.; Lee, Y. J. [Ajou Univ., Suwon (Korea, Republic of)

    2001-03-15

    The objective of project is to derive the policy recommendations to improve the efficiency of CANDU plants regulation. These policy recommendations will eventually contribute to the upgrading of Korean nuclear regulatory system and safety enhancement. During the first phase of this 2 years study, following research activities were done. On-site survey and analysis on CANDU plants regulation. Review on CANDU plants regulating experiences and current constraints. Review and analysis on the new Canadian regulatory approach.

  14. Numerical study of water management in the air flow channel of a PEM fuel cell cathode

    Energy Technology Data Exchange (ETDEWEB)

    Quan, Peng; Lai, Ming-Chia [Department of Mechanical Engineering, Wayne State University, Detroit, MI 48202 (United States)

    2007-01-10

    The water management in the air flow channel of a proton exchange membrane (PEM) fuel cell cathode is numerically investigated using the FLUENT software package. By enabling the volume of fraction (VOF) model, the air-water two-phase flow can be simulated under different operating conditions. The effects of channel surface hydrophilicity, channel geometry, and air inlet velocity on water behavior, water content inside the channel, and two-phase pressure drop are discussed in detail. The results of the quasi-steady-state simulations show that: (1) the hydrophilicity of reactant flow channel surface is critical for water management in order to facilitate water transport along channel surfaces or edges; (2) hydrophilic surfaces also increase pressure drop due to liquid water spreading; (3) a sharp corner channel design could benefit water management because it facilitates water accumulation and provides paths for water transport along channel surface opposite to gas diffusion layer; (4) the two-phase pressure drop inside the air flow channel increases almost linearly with increasing air inlet velocity. (author)

  15. Effects of cooling channel blockage on fuel plate temperature in Tehran Research Reactor

    Institute of Scientific and Technical Information of China (English)

    TABBAKH Farshid

    2009-01-01

    In this study, the variation of the temperature distribution of the fuel plate in Tehran Research Reactor core was studied in case of coolant channels blockage. While the experimental method is not possible, both the analytical and simulation methods were used to obtain the more reliable data. The results show that one channel blockage will increase the fuel temperature to about 100%, but it does not lead to clad melt down still. With further calculation and simulation it is understood that if the coolant velocity drops to 90% of its nominal value, it may causes the clad melt-ing down. At least two channels with complete blockage even at the positions far from the core center can also melt down the clad.

  16. Multiphysics Modeling of a Single Channel in a Nuclear Thermal Propulsion Grooved Ring Fuel Element

    Science.gov (United States)

    Kim, Tony; Emrich, William J., Jr.; Barkett, Laura A.; Mathias, Adam D.; Cassibry, Jason T.

    2013-01-01

    In the past, fuel rods have been used in nuclear propulsion applications. A new fuel element concept that reduces weight and increases efficiency uses a stack of grooved discs. Each fuel element is a flat disc with a hole on the interior and grooves across the top. Many grooved ring fuel elements for use in nuclear thermal propulsion systems have been modeled, and a single flow channel for each design has been analyzed. For increased efficiency, a fuel element with a higher surface-area-to-volume ratio is ideal. When grooves are shallower, i.e., they have a lower surface area, the results show that the exit temperature is higher. By coupling the physics of turbulence with those of heat transfer, the effects on the cooler gas flowing through the grooves of the thermally excited solid can be predicted. Parametric studies were done to show how a pressure drop across the axial length of the channels will affect the exit temperatures of the gas. Geometric optimization was done to show the behaviors that result from the manipulation of various parameters. Temperature profiles of the solid and gas showed that more structural optimization is needed to produce the desired results. Keywords: Nuclear Thermal Propulsion, Fuel Element, Heat Transfer, Computational Fluid Dynamics, Coupled Physics Computations, Finite Element Analysis

  17. Optimisation de la gestion du combustible dans les reacteurs CANDU refroidis a l'eau legere

    Science.gov (United States)

    Chambon, Richard

    This research has two main goals. First, we wanted to introduce optimization tools in the diffusion code DONJON, mostly for fuel management. The second objective is more practical. The optimization capabilities are applied to the fuel management problem for different CANDU reactors at refueling equilibrium state. Two kinds of approaches are considered and implemented in this study to solve optimization problems in the code DONJON. The first methods are based on gradients and on the quasi-linear mathematical programming. The method initially developed in the code OPTEX is implemented as a reference approach for the gradient based methods. However, this approach has a major drawback. Indeed, the starting point has to be a feasible point. Then, several approaches have been developed to be more general and not limited by the initial point choice. Among the different methods we developed, two were found to be very efficient: the multi-step method and the mixte method. The second kind of approach are the meta-heuristic methods. We implemented the tabu search method. Initially, it was designed to optimize combinatory variable problems. However, we successfully used it for continuous variables. The major advantage of the tabu method over the gradient methods is the capability to exit from local minima. Optimisation of the average exit burnup has been performed for CANDU-6 and ACR-700 reactors. The fresh fuel enrichment has also been optimized for ACR-700. Results match very well what the reactor physics can predict. Moreover, a comparison of the two totally different types of optimization methods validated the results we obtained.

  18. Update on use of AECL's MACSTOR module at CANDU 6 stations

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, R.; Moussalam, G.; Kachef, I. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada)

    2006-07-01

    AECL has contributed to the technology development and implementation of dry spent fuel management facilities in Canada and internationally over the last three decades. During that period, AECL has designed a number of concrete canister models and the MACSTOR module; a medium size air-cooled vault. AECL's dry storage technology was used in Canada, Korea and Romania for the construction of eight large-scale above ground dry storage facilities for CANDU spent fuel. These projects add up to a constructed capacity in excess of 5,000 MgU, that represents a significant share of the total worldwide dry storage capacity. This paper describes basic research and technology developments made at AECL's facilities to develop those dry storage technologies for its own reactors and for the operating CANDU 6 reactors. The current operating status of the facilities using concrete canisters is provided. A description of the MACSTOR 200 modules each having a capacity of 228 MgU that is in use at the Gentilly 2 and Cernavoda stations is provided. The Cernavoda spent fuel management facility was commissioned in 2003. The organisational, licensing, equipment supply and construction aspects that were necessary to deliver this turnkey project by AECL and its Romanian partners in 25 months are described. The paper also provides an outline of the joint program between AECL and KHNP-NETEC to develop the new MACSTOR/KN-400 and provides a description of this module having a capacity of 456 MgU (thus twice the MACSTOR 200 capacity) to be deployed by 2007 at the Wolsong site in Korea. (author)

  19. Numerical Investigation of Channel Geometry on the Performance of a Pem Fuel Cell

    Science.gov (United States)

    Khazaee, I.; Mohammadiun, M.

    2013-03-01

    A complete three-dimensional and single phase model for proton exchange membrane (PEM) fuel cells was used to investigate the effect of using different channels geometry on the performances, current density and gas concentration. The proposed model was a full cell model, which includes all the parts of the PEM fuel cell, flow channels, gas diffusion electrodes, catalyst layers and the membrane. Coupled transport and electrochemical kinetics equations were solved in a single domain; therefore no interfacial boundary condition was required at the internal boundaries between cell components. This computational fluid dynamics code was employed as the direct problem solver, which was used to simulate the three-dimensional mass, momentum, energy and species transport phenomena as well as the electron- and proton-transfer process taking place in a PEMFC. The results showed that the predicted polarization curves by using this model were in good agreement with the experimental results and a high performance was observed by using circle geometry for the channels of anode and cathode sides. Also, the results showed that the performance of the fuel cell improved when a rectangular channel was used.

  20. Experimental investigation on circumferential and axial temperature gradient over fuel channel under LOCA

    Science.gov (United States)

    Yadav, Ashwini Kumar; kumar, Ravi; Gupta, Akhilesh; Chatterjee, Barun; Mukhopadhyay, Deb; Lele, H. G.

    2014-06-01

    In a nuclear reactor temperature rises drastically in fuel channels under loss of coolant accident due to failure of primary heat transportation system. Present investigation has been carried out to capture circumferential and axial temperature gradients during fully and partially voiding conditions in a fuel channel using 19 pin fuel element simulator. A series of experiments were carried out by supplying power to outer, middle and center rods of 19 pin fuel simulator in ratio of 1.4:1.1:1. The temperature at upper periphery of pressure tube (PT) was slightly higher than at bottom due to increase in local equivalent thermal conductivity from top to bottom of PT. To simulate fully voided conditions PT was pressurized at 2.0 MPa pressure with 17.5 kW power injection. Ballooning initiated from center and then propagates towards the ends and hence axial temperature difference has been observed along the length of PT. For asymmetric heating, upper eight rods of fuel simulator were activated and temperature difference up-to 250 °C has been observed from top to bottom periphery of PT. Such situation creates steep circumferential temperature gradient over PT and could lead to breaching of PT under high pressure.

  1. Improvements and adaptive changes to the fuel channel fitness-for-service assessment process

    Energy Technology Data Exchange (ETDEWEB)

    Sedran, P.J. [Canadian Power Utility Services Limited, Toronto, Ontario (Canada)

    2011-07-01

    The first formal Fitness-for-Service (FFS) assessment methodology in the CANDU industry was issued to the AECB in 1991 in the CANDU Pressure Tube Fitness-for-Service Guidelines (FFSG), which were later incorporated into CSA N285.8 in the mid 1990s. While the utilities have continued to benefit greatly from repeated, successful FFS assessments, industry changes since 1991 have conspired to apply mounting pressures on the FFS community, to the potential detriment of the assessment process. This paper identifies inherent challenges, historical challenges, and more recent difficulties encountered by the FFS assessment community and gives recommendations for relieving some of the mounting pressures on the FFS assessors and for improving the FFS assessment process. (author)

  2. Development of TUF-ELOCA - a software tool for integrated single-channel thermal-hydraulic and fuel element analyses

    Energy Technology Data Exchange (ETDEWEB)

    Popescu, A.I.; Wu, E.; Yousef, W.W.; Pascoe, J. [Nuclear Safety Solutions Ltd., Toronto, Ontario (Canada); Parlatan, Y. [Ontario Power Generation, Toronto, Ontario (Canada); Kwee, M. [Bruce Power, Tiverton, Ontario (Canada)

    2006-07-01

    The TUF-ELOCA tool couples the TUF and ELOCA codes to enable an integrated thermal-hydraulic and fuel element analysis for a single channel during transient conditions. The coupled architecture is based on TUF as the parent process controlling multiple ELOCA executions that simulate the fuel elements behaviour and is scalable to different fuel channel designs. The coupling ensures a proper feedback between the coolant conditions and fuel elements response, eliminates model duplications, and constitutes an improvement from the prediction accuracy point of view. The communication interfaces are based on PVM and allow parallelization of the fuel element simulations. Developmental testing results are presented showing realistic predictions for the fuel channel behaviour during a transient. (author)

  3. Numerical study of cell performance and local transport phenomena in PEM fuel cells with various flow channel area ratios

    Science.gov (United States)

    Wang, Xiao-Dong; Duan, Yuan-Yuan; Yan, Wei-Mon

    Three-dimensional models of proton exchange membrane fuel cells (PEMFCs) with parallel and interdigitated flow channel designs were developed including the effects of liquid water formation on the reactant gas transport. The models were used to investigate the effects of the flow channel area ratio and the cathode flow rate on the cell performance and local transport characteristics. The results reveal that at high operating voltages, the cell performance is independent of the flow channel designs and operating parameters, while at low operating voltages, both significantly affect cell performance. For the parallel flow channel design, as the flow channel area ratio increases the cell performance improves because fuel is transported into the diffusion layer and the catalyst layer mainly by diffusion. A larger flow channel area ratio increases the contact area between the fuel and the diffusion layer, which allows more fuel to directly diffuse into the porous layers to participate in the electrochemical reaction which enhances the reaction rates. For the interdigitated flow channel design, the baffle forces more fuel to enter the cell and participate in the electrochemical reaction, so the flow channel area ratio has less effect. Forced convection not only increases the fuel transport rates but also enhances the liquid water removal, thus interdigitated flow channel design has higher performance than the parallel flow channel design. The optimal performance for the interdigitated flow channel design occurs for a flow channel area ratio of 0.4. The cell performance also improves as the cathode flow rate increases. The effects of the flow channel area ratio and the cathode flow rate on cell performance are analyzed based on the local current densities, oxygen flow rates and liquid water concentrations inside the cell.

  4. Numerical study of cell performance and local transport phenomena in PEM fuel cells with various flow channel area ratios

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xiao-Dong [Department of Thermal Engineering, School of Mechanical Engineering, University of Science and Technology Beijing, Beijing 100083 (China); Duan, Yuan-Yuan [Key Laboratory of Thermal Science and Power Engineering of MOE, Tsinghua University, Beijing 100084 (China); Yan, Wei-Mon [Department of Mechatronic Engineering, Huafan University, Shih-Ting 22305 (China)

    2007-10-11

    Three-dimensional models of proton exchange membrane fuel cells (PEMFCs) with parallel and interdigitated flow channel designs were developed including the effects of liquid water formation on the reactant gas transport. The models were used to investigate the effects of the flow channel area ratio and the cathode flow rate on the cell performance and local transport characteristics. The results reveal that at high operating voltages, the cell performance is independent of the flow channel designs and operating parameters, while at low operating voltages, both significantly affect cell performance. For the parallel flow channel design, as the flow channel area ratio increases the cell performance improves because fuel is transported into the diffusion layer and the catalyst layer mainly by diffusion. A larger flow channel area ratio increases the contact area between the fuel and the diffusion layer, which allows more fuel to directly diffuse into the porous layers to participate in the electrochemical reaction which enhances the reaction rates. For the interdigitated flow channel design, the baffle forces more fuel to enter the cell and participate in the electrochemical reaction, so the flow channel area ratio has less effect. Forced convection not only increases the fuel transport rates but also enhances the liquid water removal, thus interdigitated flow channel design has higher performance than the parallel flow channel design. The optimal performance for the interdigitated flow channel design occurs for a flow channel area ratio of 0.4. The cell performance also improves as the cathode flow rate increases. The effects of the flow channel area ratio and the cathode flow rate on cell performance are analyzed based on the local current densities, oxygen flow rates and liquid water concentrations inside the cell. (author)

  5. Development of CANDU ECCS performance evaluation methodology and guides

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-03-15

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

  6. Numerical Investigation of the Water Droplet Transport in a PEM Fuel Cell with Serpentine Flow Channel

    Directory of Open Access Journals (Sweden)

    Bittagopal Mondal

    2016-01-01

    Full Text Available The serpentine flow channel can be considered as one of the most common and practical channel layouts for a polymer electrolyte membrane fuel cell (PEMFC since it ensures an effective and efficient removal of water produced in a cell with acceptable parasitic load. Water management is one of the key issues to improve the cell performance since at low operating temperatures in PEMFC, water vapor condensation starts easily and accumulates the liquid water droplet within the flow channels, thus affecting the chemical reactions and reducing the fuel cell performance. In this article, a comprehensive three dimensional numerical simulation is carried out to understand the water droplet mobility in a serpentine gas flow channel for a wide range of surface properties, inlet air velocities, droplet positions (center or off-center, bottom or top and droplet sizes by deploying a finite volume based methodology. The liquid-gas interface is tracked following the volume-of-fluid (VOF method. The droplet transport is found to be greatly influenced by the surface wettability properties, inlet velocities, number of droplets emerged and initial droplet positions. Super hydrophobic surface property is not always preferable for designing the gas flow channels. It depends upon the inlet velocity conditions, droplet positions, number of droplets and surface properties.

  7. Experimental study on the performance of PEM fuel cells with interdigitated flow channels

    Science.gov (United States)

    Yan, Wei-Mon; Mei, Sheng-Chin; Soong, Chyi-Yeou; Liu, Zhong-Sheng; Song, Datong

    In this work, the effects of interdigitated flow channel design on the cell performance of proton exchange membrane fuel cells (PEMFCs) are investigated experimentally. To compare the effectiveness of the interdigitated flow field, the performance of the PEM fuel cells with traditional flow channel design is also tested. Besides, the effects of the flow area ratio and the baffle-blocked position of the interdigitated flow field are examined in details. The experimental results indicate that the cell performance can be enhanced with an increase in the inlet flow rate and cathode humidification temperature. Either with oxygen or air as the cathode fuel, the cells with interdigitated flow fields have better performance than conventional ones. With air as the cathode fuel, the measurements show that the interdigitated flow field results in a larger limiting current density, and the power output is about 1.4 times that with the conventional flow field. The results also show that the cell performance of the interdigitated flow field with flow area ratio of 40.23% or 50.75% is better than that with 66.75%.

  8. Experimental study on the performance of PEM fuel cells with interdigitated flow channels

    Energy Technology Data Exchange (ETDEWEB)

    Yan, Wei-Mon; Mei, Sheng-Chin [Department of Mechatronic Engineering, Huafan University, Shih-Ting, Taipei 223, Taiwan (ROC); Soong, Chyi-Yeou [Department of Aerospace and System Engineering, Feng Chia University, Seatwen, Taichung 40724, Taiwan (ROC); Liu, Zhong-Sheng; Song, Datong [Institute for Fuel Cell Innovation, National Research Council, 3250 East Mall, Vancouver, BC (Canada V6T 1W5)

    2006-09-29

    In this work, the effects of interdigitated flow channel design on the cell performance of proton exchange membrane fuel cells (PEMFCs) are investigated experimentally. To compare the effectiveness of the interdigitated flow field, the performance of the PEM fuel cells with traditional flow channel design is also tested. Besides, the effects of the flow area ratio and the baffle-blocked position of the interdigitated flow field are examined in details. The experimental results indicate that the cell performance can be enhanced with an increase in the inlet flow rate and cathode humidification temperature. Either with oxygen or air as the cathode fuel, the cells with interdigitated flow fields have better performance than conventional ones. With air as the cathode fuel, the measurements show that the interdigitated flow field results in a larger limiting current density, and the power output is about 1.4 times that with the conventional flow field. The results also show that the cell performance of the interdigitated flow field with flow area ratio of 40.23% or 50.75% is better than that with 66.75%. (author)

  9. Modeling and simulation of PEM fuel cell's flow channels using CFD techniques

    Energy Technology Data Exchange (ETDEWEB)

    Cunha, Edgar F.; Andrade, Alexandre B.; Robalinho, Eric; Bejarano, Martha L.M.; Linardi, Marcelo [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: efcunha@ipen.br; abodart@ipen.br; eric@ipen.br; mmora@ipen.br; mlinardi@ipen.br; Cekinski, Efraim [Instituto de Pesquisas Tecnologicas (IPT-SP), Sao Paulo, SP (Brazil)]. E-mail: cekinski@ipt.br

    2007-07-01

    Fuel cells are one of the most important devices to obtain electrical energy from hydrogen. The Proton Exchange Membrane Fuel Cell (PEMFC) consists of two important parts: the Membrane Electrode Assembly (MEA), where the reactions occur, and the flow field plates. The plates have many functions in a fuel cell: distribute reactant gases (hydrogen and air or oxygen), conduct electrical current, remove heat and water from the electrodes and make the cell robust. The cost of the bipolar plates corresponds up to 45% of the total stack costs. The Computational Fluid Dynamic (CFD) is a very useful tool to simulate hydrogen and oxygen gases flow channels, to reduce the costs of bipolar plates production and to optimize mass transport. Two types of flow channels were studied. The first type was a commercial plate by ELECTROCELL and the other was entirely projected at Programa de Celula a Combustivel (IPEN/CNEN-SP) and the experimental data were compared with modelling results. Optimum values for each set of variables were obtained and the models verification was carried out in order to show the feasibility of this technique to improve fuel cell efficiency. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-05-15

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

  11. Improvement of performance of gas flow channel in PEM fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Kuo, Jenn-Kun [Graduate Institute of Greenergy Technology, National University of Tainan, 700 Taiwan (China); Yen, Tzu-Shuang; Chen, Cha' o-Kuang [Department of Mechanical Engineering, National Cheng Kung University, Tainan, 70101 Taiwan (China)

    2008-10-15

    This study performs numerical simulations to evaluate the convective heat transfer performance and velocity flow characteristics of the gas flow channel design to enhance the performance of proton exchange membrane fuel cells (PEMFCs). To restrict the current simulations to two-dimensional incompressible flows, the flow regime is assumed to be laminar with a low Reynolds number of approximately 200. In addition, the field synergy principle is applied to demonstrate that an increased interruption within the fluid flow reduces the intersection angle between the velocity vector and the temperature gradient. The interruption within the fluid flow is induced by different type of obstacles: wave like, trapezoid like and ladder like forms and the straight form of the gas flow channel. The numerical results show that, compared to a conventional straight gas flow channel, the wave like, trapezoid like and ladder like geometry of the proposed gas flow channel increases the mean Nusselt number by a factor of approximately two. Furthermore, the periodic three patterns (wave like, trapezoid like and ladder like) structure increases the gas flow velocity in the channel and, hence, improves the catalysis reaction performance in the catalyst layer. Finally, the results show that the three patterns geometry of the gas flow channel reduces the included angle between the velocity vector and the temperature gradient. Hence, the present numerical results are consistent with the field synergy principle, which states that the convective heat transfer is enhanced when the velocity vector and temperature gradient are closely aligned with one another. (author)

  12. Quality assurance in CANDU-type fuel manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Afifi, Y.K.; El-Hakim, E. [Atomic Energy Authority, Metallurgy Dept., Nuclear Research Center, Cairo (Egypt)

    1997-07-01

    This paper is concerned with fabrication of UO{sub 2} pellets from UO{sub 2} powder by powder metallurgical methods (pelletizing and sintering). The effect of ejection to molding force ratio on the fired pellets properties was studied. It is observed that pellets cold-pressed at ratio less than 75% is cracked to two parts. The effect of the sintering temperature on the fired pellets properties was studied. It was found that the sintering of UO{sub 2} pellets at 1650 deg C leads to production of pelletswithin the qualification requirements. The data and information available in the ASTM for each step in UO{sub 2} pellets fabrication process and the technical experience (gained or published) are transformed into a group of logic flow charts (LFC'S). These logic flow charts are collected to form a module of a software to qualify the sintered pellets and also gives a technical assistance according to the ASTM for each step in the fabrication process. (author)

  13. Water behavior in a U-shaped flow channel of PEM fuel cells. Paper no. IGEC-1-036

    Energy Technology Data Exchange (ETDEWEB)

    Quan, P.; Zhou, B.; Sobiesiak, A. [Univ. of Windsor, Dept. of Mechanical, Automotive and Materials Engineering, Windsor, Ontario (Canada)]. E-mail: bzhou@uwindsor.ca; Liu, Z.-S. [National Research Council Canada, Inst. for Fuel Cell innovation, Vancouver, British Columbia (Canada)

    2005-07-01

    The behavior of liquid water inside a U-shaped channel for a Proton Exchange Membrane (PEM) fuel cell was investigated through CFD modeling with the FLUENT software package. The Volume-Of-Fluid (VOF) model was adopted to track dynamic air-water interfaces. Three cases with varying initial water phase distributions corresponding to different fuel cell operating conditions were numerically simulated to obtain a better understanding of water behavior inside the serpentine channel. The results show that the bend area in the serpentine flow channel has significant effects on the flow field, which in turn affects the air-water flow and water liquid distribution inside the channel or along the interior channel surfaces, thus affecting fuel cell performance. (author)

  14. Studies on supercritical water reactor fuel assemblies using the sub-channel code COBRA-EN

    Energy Technology Data Exchange (ETDEWEB)

    Ammirabile, Luca, E-mail: luca.ammirabile@ec.europa.e [European Commission, JRC, Institute for Energy, Westerduinweg 3, 1755 LE Petten (Netherlands)

    2010-10-15

    In the Generation IV International Forum (GIF) program, the supercritical water reactor (SCWR) concept is among the six innovative reactor types selected for development in the near future. In principle the higher efficiency and better economics make the SCWR concept competitive with the current reactor design. Due to different technical challenges that, however exist, fuel assembly design represents a crucial aspect for the success of this concept. In particular large density variations, low moderation, heat transfer enhancement and deterioration have a strong effect on the core design parameters. Only a few computational tools are currently able to perform sub-channel thermal-hydraulic analysis under supercritical water conditions. At JRC-IE the existing sub-channel code COBRA-EN has been improved to work above the critical pressure of water. The water properties package of the IAPWS Industrial Formulation 1997 was integrated in COBRA-EN to compute the Thermodynamic Properties of Water and Steam. New heat transfer and pressure drop correlations more indicated for the supercritical region of water have also been incorporated in the code. As part of the efforts to appraise the new code capabilities, a code assessment was carried out on the hexagonal fuel assembly of a fast supercritical water reactor. COBRA-EN was also applied in combination with the neutronic code MCNP to investigate on the use of hydride fuel in the HPLWR supercritical water fuel assembly. The results showed that COBRA-EN was able to reproduce the results of similar studies with acceptable accuracy. Future activities will focus on the validation of the code against experimental data and the implementation of new features (counter-current moderator channel, wall, and wire-wrap models).

  15. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Young; Park, Kun Chul [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2003-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SDS-1, SDS2, ECCS, and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  16. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong; Lee, Jae Young; Bang, Kwang Hyun [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2001-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SOS-1, SOS-2, ECCS and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  17. Performance of pressure tubes in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

  18. Experimental study on the two phase flow behavior in PEM fuel cell parallel channels with porous media inserts

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Jixin [Mechanical and Aerospace Engineering, University of California, Irvine, Irvine, CA 92697-3975 (United States)

    2010-02-15

    In this study, the air-water two phase flow behavior in PEM fuel cell parallel channels with porous media inserts was experimentally investigated using a self-designed and manufactured transparent assembly. The visualization images of the two phase flow in channels with porous media inserts were presented and three patterns were summarized. Compared with the traditional hollow channel design, the novel configuration featured less severe two phase flow mal-distribution and self-adjustment to water amount in channels, although a higher pressure drop was introduced due to the porous media inserts. The dominant frequency of pressure drop signal was found to be a diagnostic tool for water behavior in channels. The novel flow channel design with porous media inserts may become a solution to the water management problem in PEM fuel cells. (author)

  19. Experimental study on the two phase flow behavior in PEM fuel cell parallel channels with porous media inserts

    Science.gov (United States)

    Chen, Jixin

    In this study, the air-water two phase flow behavior in PEM fuel cell parallel channels with porous media inserts was experimentally investigated using a self-designed and manufactured transparent assembly. The visualization images of the two phase flow in channels with porous media inserts were presented and three patterns were summarized. Compared with the traditional hollow channel design, the novel configuration featured less severe two phase flow mal-distribution and self-adjustment to water amount in channels, although a higher pressure drop was introduced due to the porous media inserts. The dominant frequency of pressure drop signal was found to be a diagnostic tool for water behavior in channels. The novel flow channel design with porous media inserts may become a solution to the water management problem in PEM fuel cells.

  20. Qualification of inspection systems in the CANDU nuclear industry

    Energy Technology Data Exchange (ETDEWEB)

    Baron, J.A. [CANDU Owners Group, CANDU Inspection Qualification Bureau, Toronto, Ontario (Canada)

    2014-01-15

    Most jurisdictions that generate electricity through nuclear-electric plants have imposed requirements on inspection systems beyond the typical Level 1, 2 and 3 found in personnel qualification/certification schemes. The paper discusses the rationale for this obligation and describes how the requirement for inspection qualification has been implemented for CANDU plants. The paper discusses the qualification structure and process, including a brief overview of experience to-date in qualifying Inspection Procedures. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

  2. Advancement of safeguards inspection technology for CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Sung; Park, W. S.; Cha, H. R.; Ham, Y. S.; Lee, Y. G.; Kim, K. P.; Hong, Y. D

    1999-04-01

    The objectives of this project are to develop both inspection technology and safeguards instruments, related to CANDU safeguards inspection, through international cooperation, so that those outcomes are to be applied in field inspections of national safeguards. Furthermore, those could contribute to the improvement of verification correctness of IAEA inspections. Considering the level of national inspection technology, it looked not possible to perform national inspections without the joint use of containment and surveillance equipment conjunction with the IAEA. In this connection, basic studies for the successful implementation of national inspections was performed, optimal structure of safeguards inspection was attained, and advancement of safeguards inspection technology was forwarded. The successful implementation of this project contributed to both the improvement of inspection technology on CANDU reactors and the implementation of national inspection to be performed according to the legal framework. In addition, it would be an opportunity to improve the ability of negotiating in equal shares in relation to the IAEA on the occasion of discussing or negotiating the safeguards issues concerned. Now that the national safeguards technology for CANDU reactors was developed, the safeguards criteria, procedure and instruments as to the other item facilities and fabrication facilities should be developed for the perfection of national inspections. It would be desirable that the recommendations proposed and concreted in this study, so as to both cope with the strengthened international safeguards and detect the undeclared nuclear activities, could be applied to national safeguards scheme. (author)

  3. Analysis of Moderator System Failure Accidents by Using New Method for Wolsong-1 CANDU 6 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Dongsik; Kim, Jonghyun; Cho, Cheonhwey [Atomic Creative Technology Co., Ltd., Daejeon (Korea, Republic of); Kim, Sungmin [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2013-05-15

    To reconfirm the safety of moderator system failure accidents, the safety analysis by using the reactor physics code, RFSP-IST, coupled with the thermal hydraulics code, CATHENA is performed additionally. In the present paper, the newly developed analysis method is briefly described and the results obtained from the moderator system failure accident simulations for Wolsong-1 CANDU 6 reactor by using the new method are summarized. The safety analysis of the moderator system failure accidents for Wolsong-1 CANDU 6 reactor was carried out by using the new code system, i. e., CATHENA and RFSP-IST, instead of the non-IST old codes, namely, SMOKIN G-2 and MODSTBOIL. The analysis results by using the new method revealed as same with the results by using the old method that the fuel integrity is warranted because the localized power peak remained well below the limits and, most importantly, the reactor operation enters into the self-shutdown mode due to the substantial loss of moderator D{sub 2}O inventory from the moderator system. In the analysis results obtained by using the old method, it was predicted that the ROP trip conditions occurred for the transient cases which are also studied in the present paper. But, in the new method, it was found that the ROP trip conditions did not occur. Consequently, in the safety analysis performed additionally by using the new method, the safety of moderator system failure accidents was reassured. In the future, the new analysis method by using the IST codes instead of the non-IST old codes for the moderator system failure accidents is strongly recommended.

  4. Investigation of bio-inspired flow channel designs for bipolar plates in proton exchange membrane fuel cells

    Science.gov (United States)

    Kloess, Jason P.; Wang, Xia; Liu, Joan; Shi, Zhongying; Guessous, Laila

    Proton exchange membrane (PEM) fuel cell performance is directly related to the flow channel design on bipolar plates. Power gains can be found by varying the type, size, or arrangement of channels. The objective of this paper is to present two new flow channel patterns: a leaf design and a lung design. These bio-inspired designs combine the advantages of the existing serpentine and interdigitated patterns with inspiration from patterns found in nature. Both numerical simulation and experimental testing have been conducted to investigate the effects of two new flow channel patterns on fuel cell performance. From the numerical simulation, it was found that there is a lower pressure drop from the inlet to outlet in the leaf or lung design than the existing serpentine or interdigitated flow patterns. The flow diffusion to the gas diffusion layer was found be to more uniform for the new flow channel patterns. A 25 cm 2 fuel cell was assembled and tested for four different flow channels: leaf, lung, serpentine and interdigitated. The polarization curve has been obtained under different operating conditions. It was found that the fuel cell with either leaf or lung design performs better than the convectional flow channel design under the same operating conditions. Both the leaf and lung design show improvements over previous designs by up to 30% in peak power density.

  5. Investigation of bio-inspired flow channel designs for bipolar plates in proton exchange membrane fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Kloess, Jason P. [Department of Mechanical Engineering, University of Michigan, Ann Arbor, MI (United States); Wang, Xia; Shi, Zhongying; Guessous, Laila [Department of Mechanical Engineering, Oakland University, Rochester, MI 48309 (United States); Liu, Joan [Department of Mechanical Engineering, Olin College of Engineering, MA (United States)

    2009-03-01

    Proton exchange membrane (PEM) fuel cell performance is directly related to the flow channel design on bipolar plates. Power gains can be found by varying the type, size, or arrangement of channels. The objective of this paper is to present two new flow channel patterns: a leaf design and a lung design. These bio-inspired designs combine the advantages of the existing serpentine and interdigitated patterns with inspiration from patterns found in nature. Both numerical simulation and experimental testing have been conducted to investigate the effects of two new flow channel patterns on fuel cell performance. From the numerical simulation, it was found that there is a lower pressure drop from the inlet to outlet in the leaf or lung design than the existing serpentine or interdigitated flow patterns. The flow diffusion to the gas diffusion layer was found be to more uniform for the new flow channel patterns. A 25 cm{sup 2} fuel cell was assembled and tested for four different flow channels: leaf, lung, serpentine and interdigitated. The polarization curve has been obtained under different operating conditions. It was found that the fuel cell with either leaf or lung design performs better than the convectional flow channel design under the same operating conditions. Both the leaf and lung design show improvements over previous designs by up to 30% in peak power density. (author)

  6. Extending Pickering Fuel Channel Life to 247,000 EFPH (2020)

    Energy Technology Data Exchange (ETDEWEB)

    Malek, I. [Ontario Power Generation, Toronto, Ontario (Canada)

    2013-07-01

    This presentation is about extending the life of PNGS Pressure tubes from 210,000 EFPH (Effective Full Power Hours) to 247,000 EFPH. The l license renewal is in 2013 (R & D defined in 2009). The unknowns are the extent of degradation in some key areas. There are regulatory requirements for fuel channel extended life. It is necessary to integrate inspection capability and do-ability in outages. This would result in safe operation and demonstrated margin to extend life with operation to 2020.

  7. Field measurements of beta ray energy spectra in CANDU nuclear generating stations

    Energy Technology Data Exchange (ETDEWEB)

    Horowitz, Y.S. (Ben-Gurion Univ. of the Negev, Beersheba (Israel). Dept. of Physics); Hirning, C.R. (Ontario Hydro, Whitby, ON (Canada)); Yuen, P.S.; Aikens, M.S. (AECL Research, Chalk River, ON (Canada). Chalk River Labs.)

    1994-01-01

    Field measurements of beta ray energy spectra have been carried out at various locations in CANDU nuclear generating stations operated by Ontario Hydro. The beta ray energy spectrometer consists of a 5 cm diameter x 2 cm thick BC-404 plastic scintillator situated behind a 100 [mu]m thick, totally depleted, silicon detector. Photon events are rejected by requiring a coincidence between the two detectors. The spectrometer is capable of measuring electron energies from 125 keV to 3.5 MeV. Beta ray energy spectra have been measured for uncontaminated and contaminated fueling machine components, fueling machine swipes and a reactor containment vault. The degree of protection afforded by various articles of protective clothing has also been investigated for the various fueling machine components. Monte Carlo calculations have been used to estimate beta factors for 100 mg.cm[sup -2] and 240 mg.cm[sup -2] LiF-TLD chips, which are used as 'skin-and 'extremity' dosemeters in the Ontario Hydro Radiation Dosimetry Programme. (Author).

  8. Study on the effect of the CANFLEX-NU fuel element bowing on the critical heat flux

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Ho Chun; Cho, Moon Sung; Jeon, Ji Su

    2001-01-01

    The effect of the CANFLEX-NU fuel element bowing on the critical heat flux is reviewed and analyzed, which is requested by KINS as the Government design licensing condition for the use of the fuel bundles in CANDU power reactors. The effect of the gap between two adjacent fuel elements on the critical heat flux and onset-of-dryout power is studied. The reduction of the width of a single inter-rod gap from its nominal size to the minimum manufacture allowance of 1 mm has a negligible effects on the thermal-hydraulic performance of the bundle for the given set of boundary conditions applied to the CANFLEX-43 element bundle in an uncrept channel. As expected, the in-reactor irradiation test results show that there are no evidence of the element bow problems on the bundle performance.

  9. The test channel groups in the Wolsong-3 reactor for the evaluation of PHWR fuel integrity in the two phase flow condition

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Ji Su; Cho, Moon Sung

    2004-07-01

    This report intends to select the test channel groups in the Wolsong-3 reactor for the evaluation of PHWR fuel integrity in the two phase flow condition. First, a detailed work schedule for fuel inspection, fuel loading, storage of the discharged fuel, and in-bay visual and dimensional examination was established on the CIGAR work plan of the Wolsong-3 reactor. According to the schedule, it is necessary to suggest the appropriate test channel groups with two phase flow before the new fuel loading. For the selection of the test channel groups, the distributions of the channel flow rate and the channel exit quality are calculated by the NUCIRC code with the monthly measured operation data of the Wolsong-3 reactor during May 2001 to December 2002. The monthly calculated data could sort out the channels with the channel exit quality greater than 0.2%, in the high power channels with the power higher than 6300 kW and the flow rate greater than 25.5 kg/s, and in the low power channels with the power higher than 4800 kW and less than 5300 kW. Based on the frequency rank of these sorted channels, O16 channel with thirty other channels in the high power region and D17 channel with ten other channels in the low power region are suggested as the test channel groups.

  10. Experimental investigation and numerical comparison of the performance of a proton exchange membrane fuel cell at different channel geometry

    Science.gov (United States)

    Khazaee, I.

    2015-08-01

    In this study, the performance of a PEM fuel cell is investigated experimentally and numerically by changing the geometry of the channels. At first an experimental setup is used and three different fuel cells with rectangular, elliptical and triangular serpentine channels are constructed. The active area of each cell is 25 cm2 that its weight is 1,300 g. The material of the gas diffusion layer is carbon clothes, the membrane is nafion 117 and the catalyst layer is a plane with 0.004 g cm-2 platinum. Then a complete three-dimensional model for fuel cell is used to investigate the effect of using this channels geometry on the performance. The proposed model is a full cell model, which includes all the parts of the PEM fuel cell, flow channels, gas diffusion electrodes, catalyst layers and the membrane. Coupled transport and electrochemical kinetics equations are solved in a single domain; therefore no interfacial boundary condition is required at the internal boundaries between cell components. The results show that the predicted polarization curves by using this model are in good agreement with the experimental results. Also the results show that when the geometry of channel is rectangular the performance of the cell is better than the triangular and elliptical channel.

  11. Effect of fuel pin ballooning on the sub-channel thermal hydraulics during small break loca for Indian PHWRS

    Energy Technology Data Exchange (ETDEWEB)

    Mukhopadhyay, D.; Behera, G.H.; Bandopadhyay, S.K.; Gupta, S.K. [Bhabha Atomic Research Centre, Div. Reactor Safety, Bombay (India)

    2001-07-01

    Effect of fuel pin ballooning on the subchannel thermal-hydraulics during a small break (0.25%) located at the Reactor Inlet Feeder (RIF) has been studied for Indian PHWRs. The break leads to a low flow situation in the affected reactor channel along with delayed reactor trip. Higher power to flow ratio in the inner subchannels in comparison to outer subchannel of a 19 pin fuel bundle causes early 2-phase condition causing the flow to by pass from the inner ones to outer ones. This causes the fuel pins to experience different temperatures. Fuel pin ballooning causes reduction in the subchannel areas and further flow redistribution takes place. The transient subchannel thermal-hydraulic conditions along the reactor channel are very much different due to the power distribution and pressure drop. (authors)

  12. Performance Characteristics of a PEM Fuel Cell with Parallel Flow Channels at Different Cathode Relative Humidity Levels

    OpenAIRE

    Sang Soon Hwang; Pil Hyong Lee

    2009-01-01

    In fuel cells flow configuration and operating conditions such as cell temperature, humidity at each electrode and stoichiometric number are very crucial for improving performance. Too many flow channels could enhance the performance but result in high parasite loss. Therefore a trade-off between pressure drop and efficiency of a fuel cell should be considered for optimum design. This work focused on numerical simulation of the effects of operating conditions, especially cathode humidity, wit...

  13. Increasing the efficiency of a portable PEM fuel cell by altering the cathode channel geometry: A numerical and experimental study

    Energy Technology Data Exchange (ETDEWEB)

    Henriques, T.; Cesar, B.; Branco, P.J. Costa [Instituto Superior Tecnico (IST), Center for Innovation in Electrical and Energy Engineering, CIE3, Av. Rovisco Pais, 1049-001, Lisboa (Portugal)

    2010-04-15

    Portable fuel cells are receiving great attention today mainly because their energy density is higher than any portable battery solution. Among other types, portable polymer electrolyte membrane (PEM) fuel cells are an established technology where research on increasing their efficiency is leading product development and manufacturing. The objective of this work was to study and evaluate the redesign of a commercial portable fuel cell, improving its efficiency. A three-dimensional model of the original PEM fuel cell with parallel plus a transversal flow channel design was developed using Comsol Multiphysics, including the effects of liquid water formation and electric current production. Using this model, the effects of different channel geometries and respective cathode flow rates on the cell's performance, including the local transport characteristics, were studied. Laboratory tests with various fuel cell stacks using the new channels structure were effectuated for an evaluation of the fuel cell's performance, showing improvements in its efficiency of up to 26.4%. (author)

  14. Development of a experimental facility for the flow regime identification of two-phase flow in the inclined mini channel

    Energy Technology Data Exchange (ETDEWEB)

    Kwark, Nam Yi; Lee, Jae Young [Handong Global University, Pohang (Korea, Republic of); Lee, Sang Gyu; Kim, Man Woong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2006-07-01

    Recently, miniaturization is one of popular issues in the engineering design. Based on the successful advance in the semi-conductor industry, the new fabrication method cultivates new devices of tiny size. Generally, these miniaturized gadgets were more efficient than the normal devices. The major physical parameters governing the phenomena are changed as the scale decreases and unexpected effects open a room of improvement. For instant, the phase change phenomena in a micro or mini channel have been known to be enhanced. For developing a compact and efficient heat exchanger, the study on the two phase flow in a miniaturized flow channel is imperative. Furthermore, the produced hydrogen will be fed into the fuel cell which needs to drain out the water generated in the reaction at the diffusion layer of electrode membrane, effective removal of vapor plug is necessary. Also, in the safety study of CANDU reactor several parts of feeder pipes have the incline angles which will affect the accuracy of the safety analysis when the flow regime map of code is ill addressed. In line with the development of the RELAP-CANDU, the present work also is aimed to be studied.

  15. In-reactor performance of pressure tubes in CANDU reactors

    Science.gov (United States)

    Rodgers, D. K.; Coleman, C. E.; Griffiths, M.; Bickel, G. A.; Theaker, J. R.; Muir, I.; Bahurmuz, A. A.; Lawrence, S. St.; Resta Levi, M.

    2008-12-01

    The pressure tubes in CANDU reactors have been operating for times up to about 25 years. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behaviour and discusses the factors controlling the behaviour of these components in currently operating CANDU reactors. The mechanical properties (such as ultimate tensile strength, UTS, and fracture toughness), and delayed-hydride-cracking properties (crack growth rate Vc, and threshold stress intensity factor, KIH) change with irradiation; the former reach a limiting value at a fluence of Pressure tubes exhibit elongation and diametral expansion. The deformation behaviour is a function of operating conditions and material properties that vary from tube-to-tube and as a function of axial location. Semi-empirical predictive models have been developed to describe the deformation response of average tubes as a function of operating conditions. For corrosion and, more importantly deuterium pickup, semi-empirical predictive models have also been developed to represent the behaviour of an average tube. The effect of material variability on corrosion behaviour is less well defined compared with other properties. Improvements in manufacturing have increased fracture resistance by minimising trace elements, especially H and Cl, and reduced variability by tightening controls on forming parameters, especially hot-working temperatures.

  16. Transient computation fluid dynamics modeling of a single proton exchange membrane fuel cell with serpentine channel

    Science.gov (United States)

    Hu, Guilin; Fan, Jianren

    The proton exchange membrane fuel cell (PEMFC) has become a promising candidate for the power source of electrical vehicles because of its low pollution, low noise and especially fast startup and transient responses at low temperatures. A transient, three-dimensional, non-isothermal and single-phase mathematical model based on computation fluid dynamics has been developed to describe the transient process and the dynamic characteristics of a PEMFC with a serpentine fluid channel. The effects of water phase change and heat transfer, as well as electrochemical kinetics and multicomponent transport on the cell performance are taken into account simultaneously in this comprehensive model. The developed model was employed to simulate a single laboratory-scale PEMFC with an electrode area about 20 cm 2. The dynamic behavior of the characteristic parameters such as reactant concentration, pressure loss, temperature on the membrane surface of cathode side and current density during start-up process were computed and are discussed in detail. Furthermore, transient responses of the fuel cell characteristics during step changes and sinusoidal changes in the stoichiometric flow ratio of the cathode inlet stream, cathode inlet stream humidity and cell voltage are also studied and analyzed and interesting undershoot/overshoot behavior of some variables was found. It was also found that the startup and transient response time of a PEM fuel cell is of the order of a second, which is similar to the simulation results predicted by most models. The result is an important guide for the optimization of PEMFC designs and dynamic operation.

  17. Study of flow channel geometry using current distribution measurement in a high temperature polymer electrolyte membrane fuel cell

    Science.gov (United States)

    Lobato, Justo; Cañizares, Pablo; Rodrigo, Manuel A.; Pinar, F. Javier; Úbeda, Diego

    To improve fuel cell design and performance, research studies supported by a wide variety of physical and electrochemical methods have to be carried out. Among the different techniques, current distribution measurement owns the desired feature that can be performed during operation, revealing information about internal phenomena when the fuel cell is working. Moreover, short durability is one of the main problems that is hindering fuel cell wide implementation and it is known to be related to current density heterogeneities over the electrode surface. A good flow channel geometry design can favor a uniform current density profile, hence hypothetically extending fuel cell life. With this, it was thought that a study on the influence of flow channel geometry on the performance of a high temperature polymer electrolyte membrane (PEM) fuel cell using current distribution measurement should be a very solid work to optimize flow field design. Results demonstrate that the 4 step serpentine and pin-type geometries distribute the reactants more effectively, obtaining a relatively flat current density map at higher current densities than parallel or interdigitated ones and yielding maximum powers up to 25% higher when using oxygen as comburent. If air is the oxidant chosen, interdigitated flow channels perform almost as well as serpentine or pin-type due to that the flow conditions are very important for this geometry.

  18. A Study of the Influence of Gas Channel Parameters on HT-PEM Fuel Cell Performance Using FEM Analysis

    Directory of Open Access Journals (Sweden)

    Ionescu Viorel

    2016-01-01

    Full Text Available Proton Exchange Membrane Fuel Cells (PEMFC are highly efficient power generators, achieving up to 50–60% conversion efficiency, even in sizes of a few kilowatts. Comsol Multiphysics, a commercial solver based on the Finite Element Method (FEM was used for developing a three dimensional model of a high temperature PEMFC that can deal with both anode and cathode flow field for examining the micro flow channel with electrochemical reaction. Cathode gas flow velocity influence on the cell performance was investigated at first. Polarization curves for three different channel widths (0.8, 1.6 and 2.4 mm and three different channel depths (1, 2 and 3 mm were computed at a cathode inlet flow velocity of 0.06 m/s. Oxygen molar concentration at cathode catalyst layer-GDL channel interface and local current density variation along the cell length were also studied for specific gas channel geometries.

  19. Analysis of transport phenomena and electrochemical reactions in a micro PEM fuel cell with serpentine gas flow channels

    Directory of Open Access Journals (Sweden)

    Maher A.R. Sadiq Al-Baghdadi

    2014-01-01

    Full Text Available Micro-fuel cells are considered as promising electrochemical power sources in portable electronic devices. The presence of microelectromechanical system (MEMS technology makes it possible to manufacture the miniaturized fuel cell systems. The majority of research on micro-scale fuel cells is aimed at micro-power applications. Performance of micro-fuel cells are closely related to many factors, such as designs and operating conditions. CFD modeling and simulation for heat and mass transport in micro PEM fuel cells are being used extensively in researches and industrial applications to gain better understanding of the fundamental processes and to optimize the micro fuel cell designs before building a prototype for engineering application. In this research, full three-dimensional, non-isothermal computational fluid dynamics model of a micro proton exchange membrane (PEM fuel cell with serpentine gas flow channels has been developed. This comprehensive model accounts for the major transport phenomena such as convective and diffusive heat and mass transfer, electrode kinetics, transport and phase-change mechanism of water, and potential fields in a micro PEM fuel cell. The model explains many interacting, complex electrochemical, and transport phenomena that cannot be studied experimentally.

  20. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a second step of the whole project, and focus to the implementation of CANDU models based on the previous study. FORTRAN 90 language have been used for the development of RELAP5.MOD3/CANDU PC version. For the convenience of the previous Workstation users, the FOTRAN 77 version has been coded also and implanted into the original RELAP5 source file. The verification of model implementation has been performed through the simple verification calculations using the CANDU version. 6 refs., 15 figs., 7 tabs. (Author)

  1. Effect of channel geometrical configuration on the pressure distribution and stress failure in a running PEM fuel cell

    Directory of Open Access Journals (Sweden)

    Maher A.R. Sadiq Al-Baghdadi

    2017-01-01

    Full Text Available Proton Exchange membrane (PEM fuel cells are still undergoing intense development, and the combination of new and optimized materials, improved product development, novel architectures, more efficient transport processes, and design optimization and integration are expected to lead to major gains in performance, efficiency, durability, reliability, manufacturability and cost-effectiveness. PEM fuel cell assembly pressure is known to cause large strains in the cell components. All components compression occurs during the assembly process of the cell, but also during fuel cell operation due to membrane swelling when absorbs water and cell materials expansion due to heat generating in catalyst layers. Additionally, the repetitive channel-rib pattern of the bipolar plates results in a highly inhomogeneous compressive load, so that while large strains are produced under the rib, the region under the channels remains approximately at its initial uncompressed state. This leads to significant spatial variations in GDL thickness and porosity distributions, as well as in electrical and thermal bulk conductivities and contact resistances (both at the ribe-GDL and membrane-GDL interfaces. These changes affect the rates of mass, charge, and heat transport through the GDL, thus impacting fuel cell performance and lifetime. In this study, computational fluid dynamics (CFD model of a PEM fuel cell has been developed to simulate the pressure distribution inside the cell, which are occurring during fuel cell assembly (bolt assembling, and membrane swelling and cell materials expansion during fuel cell running due to the changes of temperature and relative humidity. The PEM fuel cell model simulated includes the following components; two bi-polar plates, two GDLs, and, an MEA (membrane plus two CLs. This model is used to study and analyse the effect of channel geometrical configuration on the mechanical behaviour of the PEM fuel cell components. The analysis

  2. Diagnostic technology for degradation of feeder pipe and fuel channel; design of nuclear fuel channel mock-up and development of signature analysis technique for its condition monitoring system

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S. K. [Inha University, Incheon (Korea)

    2002-04-01

    The main purpose of this project is to design the nuclear fuel channel and to develope the signal analysis method for the nonstationary signal. The design of mock-up is necessary for the pre-test of the nuclear plant. The fault signal due to the damage of the nuclear plant is generally nonstationary. It is difficult to analysis this non-stationary signal using traditional method. Thus in this research, time-frequency analysis, and wavelet transform are studied for the analysis of nonstationary signal. Basic program for analysis of nonstationary signal embedded in the background noise is developed. Results from this research can be applied to the early detection of damage of nuclear fuel channel. However, in order to apply to the real nuclear plant, further research project should be processed through the mock-up test. Then that result can be applied to the condition monitoring of a real nuclear plant. 10 refs., 18 figs., 2 tabs. (Author)

  3. A New, Scalable and Low Cost Multi-Channel Monitoring System for Polymer Electrolyte Fuel Cells.

    Science.gov (United States)

    Calderón, Antonio José; González, Isaías; Calderón, Manuel; Segura, Francisca; Andújar, José Manuel

    2016-03-09

    In this work a new, scalable and low cost multi-channel monitoring system for Polymer Electrolyte Fuel Cells (PEFCs) has been designed, constructed and experimentally validated. This developed monitoring system performs non-intrusive voltage measurement of each individual cell of a PEFC stack and it is scalable, in the sense that it is capable to carry out measurements in stacks from 1 to 120 cells (from watts to kilowatts). The developed system comprises two main subsystems: hardware devoted to data acquisition (DAQ) and software devoted to real-time monitoring. The DAQ subsystem is based on the low-cost open-source platform Arduino and the real-time monitoring subsystem has been developed using the high-level graphical language NI LabVIEW. Such integration can be considered a novelty in scientific literature for PEFC monitoring systems. An original amplifying and multiplexing board has been designed to increase the Arduino input port availability. Data storage and real-time monitoring have been performed with an easy-to-use interface. Graphical and numerical visualization allows a continuous tracking of cell voltage. Scalability, flexibility, easy-to-use, versatility and low cost are the main features of the proposed approach. The system is described and experimental results are presented. These results demonstrate its suitability to monitor the voltage in a PEFC at cell level.

  4. New design of a cathode flow-field with a sub-channel to improve the polymer electrolyte membrane fuel cell performance

    Science.gov (United States)

    Wang, Yulin; Yue, Like; Wang, Shixue

    2017-03-01

    The cathode flow-field design of polymer electrolyte membrane (PEM) fuel cells determines the distribution of reactant gases and the removal of liquid water. A suitable design can result in perfect water management and thus high cell performance. In this paper, a new design for a cathode flow-field with a sub-channel was proposed and had been experimentally analyzed in a parallel flow-field PEM fuel cell. Three sub-channel inlets were placed along the cathode channel. The main-channel inlet was fed with moist air to humidify the membrane and maintain high proton conductivity, whereas, the sub-channel inlet was fed with dry air to enhance water removal in the flow channel. The experimental results indicated that the sub-channel design can decrease the pressure drop in the flow channel, and the sub-channels inlet positions (SIP, where the sub-channel inlets were placed along the cathode channel) and flow rates (SFR, percentage of air from the sub-channel inlet in the total cathode flow rate) had a considerable impact on water removal and cell performance. A proposed design that combines the SIP and SFR can effectively eliminate water from the fuel cell, increasing the maximum power density by more than 13.2% compared to the conventional design.

  5. Etude de l'influence du champ magnetique dans une section d'essais thermohydraulique d'un canal de reacteur nucleaire CANDU 6

    Science.gov (United States)

    Landry-Lavoie, Renaud

    This memoir deals with the effects of the magnetic fields present in a thermal hydraulic test section of the Canadian nuclear industry. This test section is used to determine the thermal hydraulic conditions that can lead to critical heat flux in a channel of a CANDU 6 nuclear reactor. To perform their series of experiments the STERN Company used strong electric currents to heat the simulation bundles with a thermal power similar to the one found in a channel of a CANDU reactor. The materials constituting the simulation channel and its supports are of ferromagnetic nature. The strong magnetic field generated by the bundles implies that they are subjected to a magnetostatic force due to the magnetization of the ferromagnetic materials. The nuclear industry wants to know if these efforts, combined with the force of gravity, are sufficient to maintain the bundles in place in the simulation channel. The question also arises whether or not the magnetic field present in the channel can affect the parameters of boiling heat transfer. To determine the magnetic field distribution in the simulation channel, we had recourse to the magnetostatic image method and the integral method of calculation of magnetization. The results of the calculations show that the magnetostatic forces exerted by the ferromagnetic elements of the test section are inferior in magnitude to the one estimated by the STERN laboratorie. We used the mechanistic model of Sullivan et al. (1964) to evaluate the possible influence of the magnetic fields on the departure diameter of the vapor bubbles. The deviation in the frequency of bubble emission was evaluated by using the correlations of Zuber et al. (1959) and Cole (1960). By introducing a magnetostatic force in the boiling model and in the correlations, we demonstrated that the magnetic field present in the STERN test section has a negligible effect on the bubble departure diameter and their emission frequency. We conclude that the conditions in the test

  6. Electrochemical reaction and performance of proton exchange membrane fuel cells with a novel cathode flow channel shape

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Mu-Sheng [Department of Mechanical Engineering, National Chiao Tung University, Hsin-Chu 30010 (Taiwan); Department of Mechanical Engineering, Nan Kai Institute of Technology, Nantou 54243 (Taiwan); Chu, Hsin-Sen [Department of Mechanical Engineering, National Chiao Tung University, Hsin-Chu 30010 (Taiwan); Chen, Cha' o-Kuang [Department of Mechanical Engineering, National Cheng Kung University, Tainan 70101 (Taiwan); Jian, Sheng-Rui [Department of Electrophysics, National Chiao Tung University, Hsin-Chu 30010 (Taiwan)

    2007-04-15

    This study focuses on the investigation of the electrochemical reaction along a novel cathode flow channel of PEM fuel cells with various shoulder/channel (S/C) ratios at the outlet port. A three-dimensional mathematical model, considering conservation principles of mass, momentum, species and electric current is employed. Local variations of important model variables such as reactant concentration and local current density are presented by contour plots to elucidate the effects of channel geometry on transport process, catalyst reaction and cell performance. The potential fields of solid and membrane phases are also resolved in the cell domain and the driving force of the electrochemical reactions - the catalyst activation overpotential - is harnessed in modeling. Numerical calculations reveal the influence of the cathode channel configuration on the local distributions of various model variables. The results also show the dependence between optimal channel configuration and cell operating condition. At a medium reaction rate, the reaction sites underneath the shoulder region generate more current than the channel region. Therefore, a convergent channel configuration with a larger S/C ratio at the outlet port develops more current because such a design facilitates the electron transport and enhances local activation overpotential. However, as the cell voltage decreases and the reaction rate increases, such configuration loses its merit gradually as the requirement for a higher reactant concentration is more important and the reaction sites underneath the channel region have a higher reaction rate. Consequently, the divergent channel configuration with a lower S/C ratio of 0.67 performs better at a cell voltage of 0.22 V. (author)

  7. Integrated evolution of the medium power CANDU{sup MD} reactors; Evolution integree des reacteurs CANDU{sup MD} de moyenne puissance

    Energy Technology Data Exchange (ETDEWEB)

    Nuzzo, F. [AECL Accelerators, Kanata, ON (Canada)

    2002-07-01

    The aim of this document is the main improvements of the CANDU reactors in the economic, safety and performance domains. The presentation proposes also other applications as the hydrogen production, the freshening of water sea and the bituminous sands exploitation. (A.L.B.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-04-01

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

  9. Macstor system for spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

    Pattantyus, P. (Atomic Energy of Canada Ltd., Montreal, PQ (Canada). Power Projects)

    1993-01-01

    In 1989, Transnuclear Inc. and AECL jointly developed the conceptual design for the Modular Aircooled Canister Storage System (Macstor) for LWR fuel. The development effort has proceeded to the completion of successful full-scale thermal testing. In 1990, AECL adapted the Macstor System approach for use with Candu fuel. The adapted design, called Canstor, has also successfully completed full-scale thermal testing, and the final system design has been completed. (author) 1 fig.

  10. Release of segregated nuclides from spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, L.H.; Tait, J.C. [Atomic Energy Canada Ltd., Pinawa, MB (Canada). Whiteshell Laboratories

    1997-10-01

    The potential release of fission and activation products from spent nuclear fuel into groundwater after container failure in the Swedish deep repository is discussed. Data from studies of fission gas release from representative Swedish BWR fuel are used to estimate the average fission gas release for the spent fuel population. Information from a variety of leaching studies on LWR and CANDU fuel are then reviewed as a basis for estimating the fraction of the inventory of key radionuclides that could be released preferentially (the Instant Release Fraction of IRF) upon failure of the fuel cladding. The uncertainties associated with these estimates are discussed. 33 refs, 6 figs, 3 tabs.

  11. Computer code applicability assessment for the advanced Candu reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wren, D.J.; Langman, V.J.; Popov, N.; Snell, V.G. [Atomic Energy of Canada Ltd (Canada)

    2004-07-01

    AECL Technologies, the 100%-owned US subsidiary of Atomic Energy of Canada Ltd. (AECL), is currently the proponents of a pre-licensing review of the Advanced Candu Reactor (ACR) with the United States Nuclear Regulatory Commission (NRC). A key focus topic for this pre-application review is the NRC acceptance of the computer codes used in the safety analysis of the ACR. These codes have been developed and their predictions compared against experimental results over extended periods of time in Canada. These codes have also undergone formal validation in the 1990's. In support of this formal validation effort AECL has developed, implemented and currently maintains a Software Quality Assurance program (SQA) to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. This paper discusses the SQA program used to develop, qualify and maintain the computer codes used in ACR safety analysis, including the current program underway to confirm the applicability of these computer codes for use in ACR safety analyses. (authors)

  12. Application of principal component analysis for the diagnosis of neutron overpower system oscillations in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nasimi, Elnara; Gabbar, Hossam A., E-mail: Hossam.gabbar@uoit.ca

    2014-04-01

    Highlights: • Diagnosis of neutron overpower protection (NOP) in CANDU reactors. • Accurate reactor detector modeling. • NOP detectors response analysis. • Statistical methods for quantitative analysis of NOP detector behavior. - Abstract: An accurate fault modeling and troubleshooting methodology is required to aid in making risk-informed decisions related to design and operational activities of current and future generation of CANDU{sup ®} designs. This paper attempts to develop an explanation for the unanticipated detector response and overall behavior phenomena using statistical methods to compliment traditional engineering analysis techniques. Principal component analysis (PCA) methodology is used for pattern recognition using a case study of Bruce B zone-control level oscillations.

  13. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L. [Whiteshell Labs., Pinawa (Canada)] [and others

    1997-04-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis.

  14. A study for good regulation of the CANDU's in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se Ki; Shin, Y. K.; Joe, S. K.; Kim, J. S.; Yu, Y. J.; Lee, Y. J. [Ajou Univ., Suwon (Korea, Republic of)

    2002-03-15

    The objective of project is to derive the policy recommendations to improve the efficiency of CANDU plants regulation. These policy recommendations will eventually contribute to the upgrading of Korean nuclear regulatory system and safety enhancement. During the second phase of this 2 years study, following research activities were done. Review the technical basis and framework of the new Canadian Regulation System and IAEA. Analysis on the interview of Wolsung operation staffs to identify important safety issues and regulation problems experienced at operation. Providing a plan of CANDU regulation system enhancement program.

  15. Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

    Science.gov (United States)

    Springborn, R. H.

    1971-01-01

    Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

  16. Numerical study of droplet dynamics in a polymer electrolyte fuel cell gas channel using an embedded Eulerian-Lagrangian approach

    Science.gov (United States)

    Jarauta, Alex; Ryzhakov, Pavel; Secanell, Marc; Waghmare, Prashant R.; Pons-Prats, Jordi

    2016-08-01

    An embedded Eulerian-Lagrangian formulation for the simulation of droplet dynamics within a polymer electrolyte fuel cell (PEFC) channel is presented. Air is modeled using an Eulerian formulation, whereas water is described with a Lagrangian framework. Using this framework, the gas-liquid interface can be accurately identified. The surface tension force is computed using the curvature defined by the boundary of the Lagrangian mesh. The method naturally accounts for material property changes across the interface and accurately represents the pressure discontinuity. A sessile drop in a horizontal surface, a sessile drop in an inclined plane and droplets in a PEFC channel are solved for as numerical examples and compared to experimental data. Numerical results are in excellent agreement with experimental data. Numerical results are also compared to results obtained with the semi-analytical model previously developed by the authors in order to discuss the limitations of the semi-analytical approach.

  17. Ultrasonic inspection of rocket fuel model using laminated transducer and multi-channel step pulser

    Science.gov (United States)

    Mihara, T.; Hamajima, T.; Tashiro, H.; Sato, A.

    2013-01-01

    For the ultrasonic inspection for the packing of solid fuel in a rocket booster, an industrial inspection is difficult. Because the signal to noise ratio in ultrasonic inspection of rocket fuel become worse due to the large attenuation even using lower frequency ultrasound. For the improvement of this problem, we tried to applied the two techniques in ultrasonic inspection, one was the step function pulser system with the super wideband frequency properties and the other was the laminated element transducer. By combining these two techniques, we developed the new ultrasonic measurement system and demonstrated the advantages in ultrasonic inspection of rocket fuel model specimen.

  18. Estimation of the coolant flow through a natural circulation BWR fuel channel applying and equivalent electrical model

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, DEPFI, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Campus Morelos en IMTA, Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: julfi_ig@yahoo.com.m [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2010-10-15

    This work presents the design and implementation of an advanced controller for a reduced order model of a BWR reactor core cooled by natural circulating water, which allows real time estimates of coolant flows through fuel assemblies about standard neutron flux strings. Nuclear power plants with boiling water reactors control individual fuel assembly coolant flows by forced circulation using external or internal water pumps and different core support plate orifices. These two elements reduce flow dependency on local channel pressure drops. In BWR reactors using only natural circulation coolant flows, these two elements are not available and therefore individual channel coolant flows are highly dependent in local conditions, such as power distributions and local pressure drops. Therefore it is expected that grater uncertainties in these variables be used during safety, fuel management and other analysis, which in turns may lead to increased operation penalties, such as tighter operating limits. The objective of this work is to asses by computer simulations means to reduce uncertainties in the measurement of fuel assembly coolant flows and eventually the associated penalties. During coolant phase transitions, pressure drops and local power may alter local natural circulation through fuel assemblies and flow estimates can be helped or not by control rod moves. This work presents the construction of an optimal controller for a core flow estimator based on a reduced order model of the coolant going though the reactor vessel components and nuclear core. This model is to be driven by plant signals from standard BWR instrumentation in order to estimate the coolant flows in selected fuel assemblies about a LPRM string. For this purpose an equivalent electrical model has been mathematically developed and numerically tested. The power-flow maps of typical BRW are used as steady state references for this equivalent model. Once these were fully reproduced for steady state

  19. Safety design of next generation SUI of CANDU stations

    Energy Technology Data Exchange (ETDEWEB)

    Nasimi, Elnara [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe St. N., Oshawa, L1H 7K4 ON (Canada); Gabbar, Hossam A., E-mail: hossam.gabbar@uoit.ca [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe St. N., Oshawa, L1H 7K4 ON (Canada)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer Review of current SUI technologies and challenges. Black-Right-Pointing-Pointer Propose a new type of SUI detectors. Black-Right-Pointing-Pointer Propose a new SUI system architecture and layout. Black-Right-Pointing-Pointer Propose implementation procedure for SUI with reduced risks. - Abstract: Due to the age and operating experience of Nuclear Power Plants, equipment ageing and obsolescence has become one of the main challenges that need to be resolved for all systems, structures and components in order to ensure a safe and reliable production of energy. This paper summarizes the research into a methodology for modernization of Start-Up Instrumentation (SUI), both in-core and Control Room equipment, using a new generation of detectors and cables in order to manage obsolescence. The main objective of this research is to develop a new systematic approach to SUI installation/replacement procedure development and optimization. Although some additional features, such as real-time data monitoring and storage/archiving solutions for SUI systems are also examined to take full advantage of today's digital technology, the objectives of this study do not include detailed parametrical studies of detector or system performance. Instead, a number of technological, operational and maintenance issues associated with Start-Up Instrumentation systems at Nuclear Power Plants (NPPs) will be identified and a structured approach for developing a replacement/installation procedure that can be standardized and used across all of the domestic CANDU (Canadian Deuterium Uranium) stations is proposed.

  20. Signal processing system design for improved shutdown system of CANDU{sup ®} nuclear reactors in large break LOCA events

    Energy Technology Data Exchange (ETDEWEB)

    Gabbar, Hossam A., E-mail: hossam.gabbar@uoit.ca [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, ON, Canada L1H 7K4 (Canada); Faculty of Engineering and Applied Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, ON, Canada L1H 7K4 (Canada); Xia, Lingzhi; Isham, Manir U. [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, ON, Canada L1H 7K4 (Canada); Ponomarev, Vladimir [Megawatt Solutions, 1235 Radom St., unit 68, Pickering, ON, Canada L1W 1J3 (Canada)

    2016-03-15

    Highlights: • Neutronic signal processing system design to improve CANDU SDS1 performance. • Reactor modeling for CANDU LLOCA transient. • MATLAB/Simulink system implementation for the SDS1 trip logic. • Increasing the SDS1 trip response. - Abstract: For CANDU reactors, several options to improve CANDU nuclear power plant operation safety margin have been investigated in this paper. A particular attention is paid to the response time of CANDU shutdown system number 1 (SDS1) in case of large break loss of coolant accident (LLOCA). Based on point kinetic method, a systematic fundamental analysis is performed to CANDU LLOCA event, and the power transient signal is generated. In order to improve the SDS1 response time during LLOCA events, an innovative power measurement and signal processing system is particularly designed. The new signal processing system is implemented with the input of the LLOCA power transient, and the simulation results of the reactor trip time and signal are compared to those of the existing system in CANDU power plants. It is demonstrated that the new signal processing system can not only achieve a shorter reactor trip time than the existing system, but also accommodate the spurious trip immunity. This will significantly enhance the safety margin for the power plant operation, or bring extra economical benefits to the power plant units.

  1. Water transport in the cathode channels of direct methanol fuel cells; Wasseraustrag aus den Kathodenkanaelen von Direkt-Methanol-Brennstoffzellen

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, Alexander

    2011-10-26

    Mass transport phenomena are vital for the operating performance of direct methanol fuel cells. In particular, the discharge of liquid water from the cathode channels is crucial for the supply of oxygen to the cathode and thus for operational stability. Droplets of water in the pores of the the diffusion layer and the cathode channels may lower the power output and induce locally negative current densities as they considerably limit the oxygen supply. This work investigates the water discharge from the cathode channels using neutron radiography, synchrotron radiography and locally resolved current density measurements and it identifies ways of improving the operational stability. Neutron radiography is a measuring technique suitable for detecting the water distribution in fuels cells under operating conditions. Synchrotron radiography is a method complementary to neutron radiography, allowing a more detailed analysis of smaller areas. Special test cells adapted to both measuring methods are developed. Their electrode areas are radiographed either frontally or laterally. To enable locally resolved current density measurements, a printed circuit board with a segmented contact area is integrated into each of the test cells. The measuring technique used is based on compensated sensor resistors, which ensure a reactionless measurement. In addition, the temperature distribution and the pressure drop on the cathod side are recorded. In order to correlated the water distribution, the current density distribution and the pressure drop, neutron radiography and synchrotron radiography are both combined with locally resolved current density measurements. Furthermore, current density measurements are performed under constant laboratory conditions to study the variation of paramenters. A measurement with a stack is also performed. The experiments reveal fundamental interdependencies between different factors and the discharge of water. At a given air ratio, the geometry and the

  2. Homogeneous kinetics and equilibrium predictions of coking propensity in the anode channels of direct oxidation solid-oxide fuel cells using dry natural gas

    Science.gov (United States)

    Walters, Kevin M.; Dean, Anthony M.; Zhu, Huayang; Kee, Robert J.

    Direct electrochemical oxidation (DECO) solid-oxide fuel cells (SOFCs) offer the potential to generate electrical power from hydrocarbon fuels without the need for upstream fuel processing, such as reforming. However, with pure hydrocarbon fuel entering the flow channels at temperatures typically above 700 °C, fuel pyrolysis can cause molecular-weight growth and the formation of deleterious carbonaceous deposits. This paper, which develops a plug-flow model for fuel (natural gas surrogate) within the anode channels, considers the elementary gas-phase chemical kinetics of fuel pyrolysis and oxidation. It also considers the limiting case of local chemical equilibrium. Formation of cyclic hydrocarbon species is used to indicate deposit propensity. Results show that the likelihood of deposit formation depends strongly on cell temperature, current density, and residence time. Generally speaking, equilibrium favors deposit formation early in the channel whereas, owing to limited residence time, the homogeneous finite-rate kinetics predicts relatively low levels of deposit precursors. In the downstream portions, because of electrochemical oxygen flux though the electrode-electrolyte membrane, chemical equilibrium shifts strongly away from deposit formation to volatile carbon-oxygen species. However, the homogeneous finite-rate kinetics predictions show a continuing increase in coking propensity.

  3. Investigation on the liquid water droplet instability in a simulated flow channel of PEM fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Tae Hun; Kim, Bok Yung; Kim, Han Sang; Min, Kyoung Doug [Seoul National University, Seoul (Korea, Republic of)

    2008-05-15

    To investigate the characteristics of water droplets on the gas diffusion layer from both top-view and side-view of the flow channel, a rig test apparatus was designed and fabricated with prism attached plate. This experimental device was used to simulate the growth of a single liquid water droplet and its transport process with various air flow velocity and channel height. Not only dry condition but also fully humidified condition was also simulated by using a water absorbing sponge. The detachment height of the water droplet with dry and wet conditions was measured and analyzed. It was found that the droplet tends towards becoming unstable by decreased channel height, increased flow velocity or making a gas diffusion layer (GDL) dryer. Also, peculiar behavior of the water droplet in the channel was presented like attachment to hydrophilic wall or sudden breaking of droplet in case of fully hydrated condition. The simplified force balance model matches with experimental data as well

  4. EVALUATION OF THE APPLICABLE REACTIVITY RANGE OF A REACTIVITY COMPUTER FOR A CANDU-6 REACTOR

    Directory of Open Access Journals (Sweden)

    EUN KI LEE

    2014-04-01

    Full Text Available Recently, a CANDU digital reactivity computer system (CDRCS to measure the worth of the liquid zone controller in a CANDU-6 was developed and successfully applied to a physics test of refurbished Wolsong Unit 1. In advance of using the CDRCS, its measureable reactivity range should be investigated and confirmed. There are two reasons for this investigation. First, the CANDU-6 has a larger reactor and smaller excore detectors than a general PWR and consequently the measured reactivity is likely to reflect the peripheral power variation only, not the whole core. The second reason is photo neutrons generated from the interaction of the moderator and gamma-rays, which are never considered in a PWR. To evaluate the limitations of the CDRCS, several tens of three-dimensional steady and transient simulations were performed. The simulated detector signals were used to obtain the dynamic reactivity. The difference between the dynamic reactivity and the static worth increases in line with the water level changes. The maximum allowable reactivity was determined to be 1.4 mk in the case of CANDU-6 by confining the difference to less than 1%.

  5. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Spellman, D.J.; Bevard, B.B. [and others

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

  6. The effect of number and configuration of sediment microbial fuel cells on their performance in an open channel architecture

    Science.gov (United States)

    Abazarian, Elham; Gheshlaghi, Reza; Mahdavi, Mahmood A.

    2016-09-01

    The aim of this study is to investigate the effect of the number of sediment microbial fuel cells (SMFCs) with different configurations on the electricity generation in open channels. For this purpose, two open channels, one with three and another with four SMFCs, were operated over a long period of time and their performances were analyzed. Individual SMFCs followed an almost similar trend for electricity generation during the operation time. In addition, it was found that three-SMFC stacking in series mode produced a relatively higher maximum power density (18 mW m-2) than four-SMFC stacking (16 mW m-2). However, in parallel mode, four-SMFC stacking had a higher maximum power density of 90 mW m-2 compared to three-SMFC stacking with maximum power density of 78 mW m-2. These findings indicate that different numbers of cells significantly affect electricity generation depending on the type of SMFC configuration in the channel. The results would be helpful for application of SMFCs in large-scale.

  7. Droplet and slug formation in polymer electrolyte membrane fuel cell flow channels: The role of interfacial forces

    Science.gov (United States)

    Colosqui, Carlos E.; Cheah, May J.; Kevrekidis, Ioannis G.; Benziger, Jay B.

    A microfluidic device is employed to emulate water droplet emergence from a porous electrode and slug formation in the gas flow channel of a PEM fuel cell. Liquid water emerges from a 50 μm pore forming a droplet; the droplet grows to span the entire cross-section of a microchannel and transitions into a slug which detaches and is swept downstream. Droplet growth, slug formation, detachment, and motion are analyzed using high-speed video images and pressure-time traces. Slug volume is controlled primarily by channel geometry, interfacial forces, and gravity. As water slugs move downstream, they leave residual micro-droplets that act as nucleation sites for the next droplet-to-slug transition. Residual liquid in the form of micro-droplets results in a significant decrease in slug volume between the very first slug formed in an initially dry channel and the ultimate "steady-state" slug. A physics-based model is presented to predict slug volumes and pressure drops for slug detachment and motion.

  8. Measurements of turbulent pressures of flow in a water-conveying pipe containing a simulation fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, F.; Cao, J.; Yu, S.D. [Ryerson Univ., Dept. of Mechanical and Industrial Engineering, Toronto, Ontario (Canada)

    2008-07-01

    A test apparatus was set up to investigate the turbulent flows and flow induced vibrations in a fluid-conveying pipe containing a CANDU 43-element simulation fuel bundle. The fuel bundle is immersed in test pipe of 4-inch in diameter. A centrifugal pump circulates fresh water with a maximum velocity of 9 m/s at full pump power. The pressure fluctuation near the inner surface of the flow channel was measured at various locations using a pressure transducer and a data acquisition system. It was found that the turbulence away from the test section containing the simulation fuel bundle is largely caused by the pipe flow of high Reynolds number; the turbulence near and inside the bundle structures is the result of pipe flow and fluid-solid interactions. The measurements of pressures near the fuel bundle structure showed that the power spectral density (PSD) of pressure fluctuation has a frequency range of 1-300 Hz, and a normalized maximum pressure range of 0.04 to 0.05 times dynamic pressure. The effects of bundle angular alignments and subchannels on the pressure spectra, Strouhal number range, and streamwise pressure drop are also investigated in this paper. Results presented in this paper are useful in validating the computational models for flow-induced fluid forces that cause the fuel bundle structure to rock and fret. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-15

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

  10. SAFIRE - a robotic inspection system for CANDU feeders

    Energy Technology Data Exchange (ETDEWEB)

    Buckingham, R. [OC Robotics, Bristol (United Kingdom)

    2011-07-01

    The condition of primary circuit feeder pipes in CANDU reactors is relevant to the commercial viability and plant life. One known wear mechanism is external fretting between feeder pipes and adjacent services or support structures, particularly within the Upper Feeder Cabinet (UFC). Fretting leads to wall thinning which must not exceed certain agreed limits. Chafe shields have been added to protect the feeder pipes. Regular inspections are required of the chafe shields, feeder pipes and other structures that may cause feeder damage. Historically, the dose received by inspectors conducting this work has been significant. For this reason Ontario Power Generation has invested in a remotely operated robot system to conduct visual inspections within the UFC. This system, called SAFIRE for 'Snake-Arm Feeder Inspection Robot Equipment' has been deployed at Pickering during 2010 and 2011 and has been used to inspect areas that are extremely difficult to inspect with existing manual techniques. The 2011 scope of work included inspection of a total of 660 feeder pipes in three UFC quadrants, in two reactors. The full scope was completed over a one-month period in Autumn 2011 in which SAFIRE was used during 23, twelve hour shifts. This included two periods each of 72 hours of continuous operation using multiple teams of operators. SAFIRE is remote controlled delivery system for multiple cameras to record still images and video. The main system elements include a snake-arm robot mounted on a mobile vehicle. It can be controlled from up to 500m away using a fibre/copper connection. The snake-arm is 2.2m long, 25mm wide and has 18 degrees of freedom. It is designed to snake between the rows of feeder pipes to inspect feeder/hanger interfaces, both above and below the feeder cabinet catwalks. Future upgrades offer the potential to add additional tools to increase functionality. This paper describes the SAFIRE development process from inception to operational experience

  11. Safety assessment for the CANFLEX-NU fuel bundles with respect to the 37-element fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Suk, H. C.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-11-01

    The KAERI and AECL have jointly developed an advanced CANDU fuel, called CANFLEX-NU fuel bundle. CANFLEX 43-element bundle has some improved features of increased operating margin and enhanced safety compared to the existing 37-element bundle. Since CANFLEX fuel bundle is designed to be compatible with the CANDU-6 reactor design, the behaviour in the thermalhydraulic system will be nearly identical with 37-element bundle. But due to different element design and linear element power distribution between the two bundles, it is expected that CANFLEX fuel behaviour would be different from the behaviour of the 37-element fuel. Therefore, safety assessments on the design basis accidents which result if fuel failures are performed. For all accidents selected, it is observed that the loading of CANFLEX bundle in an existing CANDU-6 reactor would not worsen the reactor safety. It is also predicted that fission product release for CANFLEX fuel bundle generally is lower than that for 37-element bundle. 3 refs., 2 figs., 2 tabs. (Author)

  12. Study on the requirement for the fuel test loop performance in the HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Chul; Kim, Hark Rho

    2000-06-01

    The requirement of the FTL (Fuel Test Loop) performance were investigated to reaffirm the technical feasibility of the FTL facility which is in consideration to install at the HANARO. LH hole in the reflector region and OR3 in the outer core region are considered as candidate sites for the IPS (In-Pile Section) for the analysis purpose. The achievable linear power at test fuel pin(s) and neutron flux levels at the cladding are analyzed in the IPS, which accommodates CANDU or PWR test fuel. The enrichment of test fuels is assumed as natural uranium for CANDU and 3.5% or 5% for PWR. The test fuel configuration is bundle or 7-pin in LH hole but 1-pin in OR site. For the CANDU test fuel, the target linear power of 60kW/m can not be achieved for all cases. For the PWR test fuel, the target linear power of 40kW/m is obtained at fuels located in the direction of the core for only the case of 5% bundle irradiation. From a sensitivity study, the linear power at test fuel is expected to increase at least 30% of the present results if the core bumup effect, optimization of the ratio of fuel-to-moderator number density, etc., are considered in the detail design. Thus, the linear power for PWR fuel is expected to reach the target value, and that for CANDU fuel will reach the target value if enriched fuel is used. The fast neutron flux at the test fuel cladding is estimated for most cases to be lower than one-third of the target value of 10{sup 1}4 n/cm{sup 2}-sec and expected not to reach the target value.

  13. Assessment of the thorium fuel cycle in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Homan, F.J.; Allen, E.J.

    1977-01-01

    A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled.

  14. Thermal-hydraulic behavior of physical quantities at critical velocities in a nuclear research reactor core channel using plate type fuel

    Directory of Open Access Journals (Sweden)

    Sidi Ali Kamel

    2012-01-01

    Full Text Available The thermal-hydraulic study presented here relates to a channel of a nuclear reactor core. This channel is defined as being the space between two fuel plates where a coolant fluid flows. The flow velocity of this coolant should not generate vibrations in fuel plates. The aim of this study is to know the distribution of the temperature in the fuel plates, in the cladding and in the coolant fluid at the critical velocities of Miller, of Wambsganss, and of Cekirge and Ural. The velocity expressions given by these authors are function of the geometry of the fuel plate, the mechanical characteristics of the fuel plate’s material and the thermal characteristics of the coolant fluid. The thermal-hydraulic study is made under steady-state; the equation set-up of the thermal problem is made according to El Wakil and to Delhaye. Once the equation set-up is validated, the three critical velocities are calculated and then used in the calculations of the different temperature profiles. The average heat flux and the critical heat flux are evaluated for each critical velocity and their ratio reported. The recommended critical velocity to be used in nuclear channel calculations is that of Wambsganss. The mathematical model used is more precise and all the physical quantities, when using this critical velocity, stay in safe margins.

  15. Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

    Directory of Open Access Journals (Sweden)

    Itamar Iliuk

    2016-01-01

    Full Text Available Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was proposed. A partial loss of flow accident was simulated to show the behavior of the model under this type of accident. The results show that the critical heat flux was detected in the central region along the axial direction of the plate when the right water channel was blocked.

  16. CATHENA Models for Wolsong 1 Channels O6 and B10

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y.H.; Lee, K.H.; Choi, H. [Korea Electric Power Research Institute, Taejon (Korea)

    2002-07-01

    This document includes CATHENA model development of slave channel O6 and B10 for Wolsong 1 trip coverage analysis in the project of ''Development of Assessment Technologies for CANDU Reactor Power and Trip Effectivness''. (author). 11 refs., 2 tabs.

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

    Science.gov (United States)

    Kim, Hyoung Tae

    2014-06-01

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

  18. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-04-15

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is first step of the whole project, thus focus to the establishment of improvement area. The study was performed by reconsideration of the previous code assessment works and investigation of AECL design analysis tools. In order to identify the thermal hydraulic phenomena for events, the whole system of CANDU plant was divided into main functional systems and subcomponents. Each phenomena was addressed to the each subcomponent. FinaIly improvement areas of model development for auditing tool were established based on the identified phenomena.

  19. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    1998-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is first step of the whole project, thus focus to the establishment of improvement area. The study was performed by reconsideration of the previous code assessment works and investigation of AECL design analysis tools. In order to identify the thermal hydraulic phenomena for events, the whole system of CANDU plant was divided into main functional systems and subcomponents. Each phenomena was addressed to the each subcomponent. Finally improvement areas of model development for auditing tool were established based on the identified phenomena. 8 refs., 21 figs., 19 tabs. (Author)

  20. A study on the interlink of CANDU safety analysis codes with development of GUI system

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. J.; Jeo, Y. J.; Park, Q. C. [Seoul National Univ., Seoul (Korea, Republic of); Kim, H. T.; Min, B. J. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    In order to improve the CANDU safety analysis code system, the interlink of containment analysis code, PRESCON2 to the system thermal hydraulics analysis code, CATHENA, has been implemented with development of the GUI system. Before the GUI development, we partly corrected two codes to optimize on the PC environment. The interlink of two codes could be executed by introducing three interlinking variables, mass flux, mixture enthalpy, and mixture specific volume. To guarantee the robustness of the codes, two codes are extremely linked by using the GUI system. The GUI system provides much of user-friendly functions and will be improved step by step. This study is expected to improve the safety assessment system and technology for CANDU NPPs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-09-01

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

  2. Implementation of an on-line monitoring system for transmitters in a CANDU nuclear power plant

    Science.gov (United States)

    Labbe, A.; Abdul-Nour, G.; Vaillancourt, R.; Komljenovic, D.

    2012-05-01

    Many transmitters (pressure, level and flow) are used in a nuclear power plant. It is necessary to calibrate them periodically to ensure that their measurements are accurate. These calibration tasks are time consuming and often contribute to worker radiation exposure. Human errors can also sometimes degrade their performance since the calibration involves intrusive techniques. More importantly, experience has shown that the majority of current calibration efforts are not necessary. These facts motivated the nuclear industry to develop new technologies for identifying drifting instruments. These technologies, well known as on-line monitoring (OLM) techniques, are non-intrusive and allow focusing the maintenance efforts on the instruments that really need a calibration. Although few OLM systems have been implemented in some PWR and BWR plants, these technologies are not commonly used and have not been permanently implemented in a CANDU plant. This paper presents the results of a research project that has been performed in a CANDU plant in order to validate the implementation of an OLM system. An application project, based on the ICMP algorithm developed by EPRI, has been carried out in order to evaluate the performance of an OLM system. The results demonstrated that the OLM system was able to detect the drift of an instrument in the majority of the studied cases. A feasibility study has also been completed and has demonstrated that the implementation of an OLM system at a CANDU nuclear power plant could be advantageous under certain conditions.

  3. Development of best estimate auditing code for CANDU thermal-hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Won Jae; Hwang, Moon Kyu; Lim, Hong Sik [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3.The study was performed by reconsideration of the previous code assessment works and phenomena identification for essential accident scenario. Improvement areas of model development for auditing tool were identified based on the code comparison and PIRT results. Nine models have been improved significantly for the analysis of LOCA and Mon LOCA event. Conceptual problem or separate effect assessment have been performed to verify the model improvement. The linking calculation with CONTAIN 2.0 has been also enabled to establish the unified auditing code system. Analysis for the CANDU plant real transient and hypothetical LOCA bas been performed using the improved version. It has been concluded that the developed version can be utilized for the auditing analysis of LOCA and non-LOCA event for the CANDU reactor. 25 refs., 84 figs., 36 tabs. (Author)

  4. Mechanical engineering utilizing advanced engineering tools for the CANDU 9 project

    Energy Technology Data Exchange (ETDEWEB)

    Nuzzo, F.; Yu, S.K.W.; Hedges, K.R. [Atomic Energy of Canada Limited (AECL), Ontario (Canada)

    1998-05-01

    To meet the increasing challenging project requirements such as cost and schedule reduction, AECL has incorporated a comprehensive suite of integrated, advanced engineering tools for CANDU project engineering and delivery. This paper provides a description of the advanced engineering tools developed and used by AECL in the pre-project engineering of the CANDU 9 product and the construction projects such as the construction of two CANDU 6 units in Qinshan, China. The advanced mechanical engineering tools described include: the Process and Instrument Diagram ; the mechanical/piping 3D models; the CADDS/piping analysis interface (PAI) tool; the pipe support design system (SDS) tool; and the powerful equipment database tool - TeddyBase. A description of the enhanced work process will also be provided. The work process improvement is a direct result of the implementation of advanced information technology and the integration of AECL tools with commercial engineering and project tools available in the market. The use of these advanced tools results in better design quality; enhanced presentation of the engineering deliverables to construction and commissioning staff; and potential support to future plant operations (Ref.1). (author). 2 refs., 3 figs.

  5. Power distribution control of CANDU reactors based on modal representation of reactor kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Xia, Lingzhi, E-mail: lxia4@uwo.ca [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Jiang, Jin, E-mail: jjiang@eng.uwo.ca [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Luxat, John C., E-mail: luxatj@mcmaster.ca [Department of Engineering Physics, McMaster University, Hamilton, Ontario L8S 4L7 (Canada)

    2014-10-15

    Highlights: • Linearization of the modal synthesis model of neutronic kinetic equations for CANDU reactors. • Validation of the linearized dynamic model through closed-loop simulations by using the reactor regulating system. • Design of a LQR state feedback controller for CANDU core power distribution control. • Comparison of the results of this new controller against those of the conventional reactor regulation system. - Abstract: Modal synthesis representation of a neutronic kinetic model for a CANDU reactor core has been utilized in the analysis and synthesis for reactor control systems. Among all the mode shapes, the fundamental mode of the power distribution, which also coincides with the desired reactor power distribution during operation, is used in the control system design. The nonlinear modal models are linearized around desired operating points. Based on the linearized model, linear quadratic regulator (LQR) control approach is used to synthesize a state feedback controller. The performance of this controller has been evaluated by using the original nonlinear models under load-following conditions. It has been demonstrated that the proposed reactor control system can produce more uniform power distribution than the traditional reactor regulation systems (RRS); in particular, it is more effective in compensating the Xenon induced transients.

  6. ATP hydrolysis is critically required for function of CaV1.3 channels in cochlear inner hair cells via fueling Ca2+ clearance.

    Science.gov (United States)

    Weiler, Simon; Krinner, Stefanie; Wong, Aaron B; Moser, Tobias; Pangršič, Tina

    2014-05-14

    Sound encoding is mediated by Ca(2+) influx-evoked release of glutamate at the ribbon synapse of inner hair cells. Here we studied the role of ATP in this process focusing on Ca(2+) current through CaV1.3 channels and Ca(2+) homeostasis in mouse inner hair cells. Patch-clamp recordings and Ca(2+) imaging demonstrate that hydrolyzable ATP is essential to maintain synaptic Ca(2+) influx in inner hair cells via fueling Ca(2+)-ATPases to avoid an increase in cytosolic [Ca(2+)] and subsequent Ca(2+)/calmodulin-dependent inactivation of CaV1.3 channels.

  7. Three-dimensional model of a 50 cm{sup 2} high temperature PEM fuel cell. Study of the flow channel geometry influence

    Energy Technology Data Exchange (ETDEWEB)

    Lobato, Justo; Canizares, Pablo; Rodrigo, Manuel A.; Pinar, F. Javier; Mena, Esperanza; Ubeda, Diego [Chemical Enginnering Department, University of Castilla-La Mancha. Enrique Costa Novella Building, Avda. Camilo Jose Cela, n 12, 13071, Ciudad Real (Spain)

    2010-06-15

    In this work, a three-dimensional half-cell model for a 50 cm{sup 2} high temperature polyelectrolyte membrane fuel cell (HTPEMFC) has been implemented in a Computational Fluid Dynamics (CFD) application. It was solved for three different flow channel geometries: 4-step serpentine, parallel and pin-type. Each geometry leads to a very well defined current density profile which indicates that current density distribution is directly linked to the way reactants are spread over the electrode surface. The model predicts that parallel flow channels present a significant lower performance probably due to the existence of preferential paths which makes the reactant gases not to be well distributed over the whole electrode surface. This results in lower output current densities when this geometry is used, especially at high oxygen demand conditions. This behavior was also detected by experimental measurement. Serpentine and pin-type flow channels were found to perform very similarly, although slightly higher limit current densities are predicted when using serpentine geometry. Inlet flow rate as well as temperature influence were also studied. The model predicts mass transfer problems and low limit current densities when the fuel cell is fed with small oxygen flow rates, whereas no differences regarding average flow rates are noticed if it is over increased. Better fuel cell performance is predicted while temperature grows as it could be expected. (author)

  8. CANFLEX fuel bundle cross-flow endurance test (test report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Deok; Chung, C. H.; Chang, S. K.; Kim, B. D.

    1997-04-01

    As part of the normal refuelling sequence of CANDU nuclear reactor, both new and irradiated bundles can be parked in the cross-flow region of the liner tubes. This situation occurs normally for a few minutes. The fuel bundle which is subjected to the cross-flow should be capable of withstanding the consequences of cross flow for normal periods, and maintain its mechanical integrity. The cross-flow endurance test was conducted for CANFLEX bundle, latest developed nuclear fuel, at CANDU-Hot Test Loop. The test was carried out during 4 hours at the inlet cross-flow region. After the test, the bundle successfully met all acceptance criteria after the 4 hours cross-flow test. (author). 2 refs., 3 tabs.

  9. Novel Flow Field with Superhydrophobic Gas Channels Prepared by One-step Solvent-induced Crystallization for Micro Direct Methanol Fuel Cell

    Institute of Scientific and Technical Information of China (English)

    Junsheng Liang; Kuanyao Liu; Shouzuo Li; Dazhi Wang; Tongqun Ren; Xiaoying Xu; Ying Luo

    2015-01-01

    The CO2-induced capillary blocking in anode flow field is one of the key adverse factors to reduce the performance of a micro-direct methanol fuel cell (lDMFC). In order to solve this problem, new polycarbonate (PC) flow field plates with nested arrangement of hydrophilic fuel channels and superhydrophobic gas channels were designed, fabricated, and tested in this work. The gas channels were treated with solvent-induced crystallization using acetone solution. The superhydrophobicity with 160? water contact angle and 2? tilting angle was obtained on the PC substrates. A dummy cell using hydrogen peroxide decomposition reaction and a test loop were separately set up to evaluate the flow fields’ performance. It was found that a 37% pressure drop decrease can be obtained in the new serpentine flow field compared with that of the conventional one. The benefit of the new flow field to remove gas bubbles was also confirmed by an in situ visualization study on the dummy cell. Results show that the auxiliary superhydrophobic gas channels can speed up the discharge of the gas bubbles from the flow field, which will in turn improve the lDMFC performance.

  10. The effect of coupled mass transport and internal reforming on modeling of solid oxide fuel cells part I: Channel-level model development and steady-state comparison

    Science.gov (United States)

    Albrecht, Kevin J.; Braun, Robert J.

    2016-02-01

    Dynamic modeling and analysis of solid oxide fuel cell systems can provide insight towards meeting transient response application requirements and enabling an expansion of the operating envelope of these high temperature systems. SOFC modeling for system studies are accomplished with channel-level interface charge transfer models, which implement dynamic conservation equations coupled with additional submodels to capture the porous media mass transport and electrochemistry of the cell. Many of these models may contain simplifications in order to decouple the mass transport, fuel reforming, and electrochemical processes enabling the use of a 1-D model. The reforming reactions distort concentration profiles of the species within the anode, where hydrogen concentration at the triple-phase boundary may be higher or lower than that of the channel altering the local Nernst potential and exchange current density. In part one of this paper series, the modeling equations for the 1-D and 'quasi' 2-D models are presented, and verified against button cell electrochemical and channel-level reforming data. Steady-state channel-level modeling results indicate a 'quasi' 2-D SOFC model predicts a more uniform temperature distribution where differences in the peak cell temperature and maximum temperature gradient are experienced. The differences are most prominent for counter-flow cell with high levels of internal reforming. The transient modeling comparison is discussed in part two of this paper series.

  11. Development of thermal-hydraulic models for the safety evaluation of CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Young; Hwang, Gi Suk; Jung, Yun Sik [Handong Univ., Pohang (Korea, Republic of); No, Hee Cheon; Moon, Young Min [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2003-03-15

    The objective of the present research is to evaluate the safety analysis for CANDU and to improve the Horizontal Stratification Entrainment Model (HSEM) of RELAP5/MOD3.3. This report includes two items : the one is the development of experimental facility for the safety evaluation of CANDU, the other is the results of comparison with the existing correlations and data. The literature reviews are performed and the database for previous off-take experiments are built. By a survey of state-of-the-articles, the deficiencies of previous works and limitations of existing models are examined. The hydraulic behavior branching through the feeder pipes from the header pipe is analyzed and the test facility of off-take experiment is designed and manufactured as the prototype CANDU6, by a proper scaling methodologies. The test facility contains various branch pipes not only for three directions (top, side and bottom), but for arbitrary directions. The experiments about the onset of entrainment and branch quality only for three directions (top, side and bottom) are carried out by using air-water as working fluids. On the whole, the existing correlations predict the present experimental results well branch quality, entrainment, which validates the availability of experimental facility and methodology. Especially, for the branch quality with top and bottom branches, the different results are shown because of the unstable flow regimes in the horizontal pipe and the different branch diameters. The deficiencies of previous works and limitations of existing models are considered. The off-take experiment for arbitrary branch angles continues as the next year research.

  12. CFD Analysis of Dry Storage System for CANDU Spent Fuel using Fluent 6.2

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Byung Soo; Jeong, Yong Hoon; Chang, Soon Heung Chang [KAIST, Taejon (Korea, Republic of)

    2006-07-01

    To obtain the licensing for MACSTOR/KN-400 developed by KHNP, 3-D CFD analysis was demanded to prove that the maximum temperature was not over the limited temperature (93 .deg. C locally and 66 .deg. C averagely for concrete). Though the thermal-hydraulic prediction by CATHENA-code show the reliable results, that could not the temperature distribution. That is, that could not predict the location of maximum temperature well. In this study, the analysis of the temperature distribution on the natural convection flow with thermal radiation shows the concrete temperature distribution. It was different from the predicted results by CATHENA-code. Therefore, to obtain the licensing for AMCSTOR/KN- 400, CFD analysis should be performed by 3-D CFD code like FLUENT at the same time.

  13. Parallel CFD simulations of turbulent flows inside a CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, F.; Yu, S.D.; Cao, J. [Ryerson Univ., Dept. of Mechanical and Industrial Engineering, Toronto, Ontario (Canada)], E-mail: fabbasia@ryerson.ca

    2008-07-01

    Large Eddy Simulation (LES) is used to study the turbulent flow inside a 43-rod bundle. The two LES models developed in this paper are of dynamic Smagorinsky type, featuring a satisfactory prediction of anisotropic turbulence intensity and frequency. The first model, by taking advantage of the geometric periodicity, deals with one seventh of a rod bundle; it is developed for studying the axial, lateral turbulence intensities and frequencies in the centers of subchannels and narrow-gap regions. The second model, dealing with the full rod bundle inside a pressure tube with nominal eccentricity, is developed for studying the turbulent fluid forces acting on the bundle. In order to accelerate the solution process for the two large CFD models, the parallelized CFD technique is utilized in connection with 24 processors. The numerical results, obtained for a test case (an eight-rod bundle), are in good agreement with those experimental data available in the literature. Numerical simulations of turbulent flow phenomena within subchannels are advantageous since true flow features are difficult or costly to reveal by experiments. (author)

  14. Development of scaling laws on thermal-hydraulic effect test facility for CANDU-6 moderator

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. Y.; Jung, Y. S.; Kim, N. S. [Handong University, Pohang (Korea, Republic of); Kim, M. W.; Kim, H. J. [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2004-07-01

    The scaling laws on thermal-hydraulic effect test facility for CANDU-6 moderator (HGU-KINS) have been investigated and manufactured. The basic laws are the satisfaction of energy conservation and dimensionless number, Ar and Re, for the similarities of thermal-hydraulic properties. And then the thermal-hydraulic scaling analyses of test facilities, SPEL(1/10 scale) and STERN(1/4 scale), have been identified by the present method. As a result, in the case of SPEL, the energy conservation is confirmed, but the similarities of Ar and the heat density are not considered. In the case of STERN, the energy conservation and the characteristics of Ar were well defined. But the similarity of the heat density is unsatisfied, either. Therefore the present method was applied with 1/8 length scale. For the performance test, CFD analysis has been accomplished by CFX5. The results of flow pattern certifications and variation of axial properties with CANDU show that the present scaling method is acceptable.

  15. The application of Plant Reliability Data Information System (PRINS) to CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, S. W.; Lim, Y. H.; Park, H. C. [Korea Hydro and Nuclear Power Co., Ltd., Naah-ri 260, Yangnam-myun, Gyeongju-si, Gyeong Buk (Korea, Republic of)

    2012-07-01

    As risk-informed applications (RIAs) are actively implanted in the nuclear industry, an issue associated with technical adequacy of Probabilistic Safety Assessment (PSA) arises in its modeling and data sourcing. In Korea, PSA for all Korean NPPs has been completed and KHNP(Korea Hydro and Nuclear Power Plant Company) developed the database called the Plant Reliability Data Information System (PRinS). It has several characteristics that distinguish it from other database system such as NPRDs (INPO,1994), PRIS (IAEA), and SRDF (EdF). This database has the function of systematic data management such as automatic data-gathering, periodic data deposition and updating, statistical analysis including Bayesian method, and trend analysis of failure rate or unavailability. In recent PSA for CANDU reactor, the component failure data of EPRI ALWR URD and Component Reliability Database were preferentially used as generic data set. The error factor for most component failure data was estimated by using the information NUREG/CR-4550 and NUREG/CR-4639. Also, annual trend analysis was performed for the functional losses of components using the statistical analysis and chart module of PRinS. Furthermore, the database has been updated regularly and maintained as a living program to reflect the current status. This paper presents the failure data analysis using PRinS which provides Bayesian analysis on main components in the CANDU reactor. (authors)

  16. Ludwig: A Training Simulator of the Safety Operation of a CANDU Reactor

    Directory of Open Access Journals (Sweden)

    Gustavo Boroni

    2011-01-01

    Full Text Available This paper presents the application Ludwig designed to train operators of a CANDU Nuclear Power Plant (NPP by means of a computer control panel that simulates the response of the evolution of the physical variables of the plant under normal transients. The model includes a close set of equations representing the principal components of a CANDU NPP plant, a nodalized primary circuit, core, pressurizer, and steam generators. The design of the application was performed using the object-oriented programming paradigm, incorporating an event-driven process to reflect the action of the human operators and the automatic control system. A comprehensive set of online graphical displays are provided giving an in-depth understanding of transient neutronic and thermal hydraulic response of the power plant. The model was validated against data from a real transient occurring in the Argentine NPP Embalse Río Tercero, showing good agreement. However, it should be stressed that the aim of the simulator is in the training of operators and engineering students.

  17. Safeguards and nonproliferation aspects of a dry fuel recycling technology

    Energy Technology Data Exchange (ETDEWEB)

    Pillay, K.K.S.

    1993-05-01

    Los Alamos National Laboratory undertook an independent assessment of the proliferation potentials and safeguardability of a dry fuel recycling technology, whereby spent pressurized-water reactor (PWR) fuels are used to fuel canadian deuterium uranium (CANDU) reactors. Objectives of this study included (1) the evaluation of presently available technologies that may be useful to safeguard technology options for dry fuel recycling (2) and identification of near-term and long-term research needs to develop process-specific safeguards requirements. The primary conclusion of this assessment is that like all other fuel cycle alternatives proposed in the past, the dry fuel recycle entails prolfferation risks and that there are no absolute technical fixes to eliminate such risks. This study further concludes that the proliferation risks of dry fuel recycling options are relatively minimal and presently known safeguards systems and technologies can be modified and/or adapted to meet the requirements of safeguarding such fuel recycle facilities.

  18. Macstor dry spent fuel storage system

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F. E. [Atomic Energy of Canada Limited, Montreal (Canada)

    1996-04-15

    AECL, a Canadian Grown Corporation established since 1952, is unique among the world's nuclear organizations. It is both supplier of research reactors and heavy water moderated CANDU power reactors as well as operator of extensive nuclear research facilities. As part of its mandate, AECL has developed products and conceptual designs for the short, intermediate and long term storage and disposal of spent nuclear fuel. AECL has also assumed leadership in the area of dry storage of spent fuel. This Canadian Crown Corporation first started to look into dry storage for the management of its spent nuclear fuel in the early 1970's. After developing silo-like structures called concrete canisters for the storage of its research reactor enriched uranium fuel, AECL went on to perfect that technology for spent CANDU natural uranium fuel. In 1989 AECL teamed up with Trans nuclear, Inc.,(TN), a US based member of the international Trans nuclear Group, to extend its dry storage technology to LWR spent fuel. This association combines AECL's expertise and many years experience in the design of spent fuel storage facilities with TN's proven capabilities of processing, transportation, storage and handling of LWR spent fuel. From the early AECL-designed unventilated concrete canisters to the advanced MACSTOR concept - Modular Air-Cooled Canister Storage - now available also for LWR fuel - dry storage is proving to be safe, economical, practical and, most of all, well accepted by the general public. AECL's experience with different fuels and circumstances has been conclusive.

  19. CFD simulations of the single-phase and two-phase coolant flow of water inside the original and modified CANDU 37-element bundles

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, F.; Hadaller, G.I.; Fortman, R.A. [Stern Laboratories, Hamilton, Ontario (Canada)

    2010-07-01

    Single-phase (inlet temperature of 180° C, outlet pressure of 9 MPa, total power of 2 MW and flow rate of 13.5 Kg/s), and two-phase (inlet temperature of 265° C, outlet pressure of 10 MPa, total power of 7.126 MW and flow rate of 19 Kg/s) water flows inside a CANDU thirty seven element fuel string are simulated using a Computational Fluid Dynamics (CFD) code with parallel processing and results are presented in this paper. The analyses have been performed for the original and modified (11.5 mm center element diameter) designs with skewed cosine axial heat flux distribution and 5.1% diametral creep of the pressure tube. The CFD results are in good agreement with the expected temperature and velocity cross-sectional distributions. The effect of the pressure tube creep on the flow bypass is detected, and the CHF improvement in the core region of the modified design is confirmed. The two-phase flow model reasonably predicted the void distribution and the role of interfacial drag on increasing the pressure drop. In all CFD models, the appendages were shown to enhance the production of cross flows and their corresponding flow mixing and asymmetry. (author)

  20. Post irradiation test report of irradiated DUPIC simulated fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Jung, I. H.; Moon, J. S. and others

    2001-12-01

    The post-irradiation examination of irradiated DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) simulated fuel in HANARO was performed at IMEF (Irradiated Material Examination Facility) in KAERI during 6 months from October 1999 to March 2000. The objectives of this post-irradiation test are i) the integrity of the capsule to be used for DUPIC fuel, ii) ensuring the irradiation requirements of DUPIC fuel at HANARO, iii) performance verification in-core behavior at HANARO of DUPIC simulated fuel, iv) establishing and improvement the data base for DUPIC fuel performance verification codes, and v) establishing the irradiation procedure in HANARO for DUPIC fuel. The post-irradiation examination performed are {gamma}-scanning, profilometry, density, hardness, observation the microstructure and fission product distribution by optical microscope and electron probe microanalyser (EPMA)

  1. Probabilistic integrity assessment of CANDU pressure tube for the consideration of flaw generation time

    Energy Technology Data Exchange (ETDEWEB)

    Kwak, Sang Log; Kim, Young Jin [Sungkyunkwan Univ., Seoul (Korea, Republic of); Lee, Joon Seong [Kyonggi Univ., Seoul (Korea, Republic of); Park, Youn Won [KINS, Taejon (Korea, Republic of)

    2001-07-01

    This paper describes a Probabilistic Fracture Mechanics (PFM) analysis based on Monte Carlo (MC) simulation. In the analysis of CANDU pressure tube, it is necessary to perform the PFM analyses based on statistical consideration of flaw generation time. A depth and an aspect ratio of initial semi-elliptical surface crack, a fracture toughness value, Delayed Hydride Cracking (DHC) velocity, and flaw generation time are assumed to be probabilistic variables. In all the analyses, degradation of fracture toughness due to neutron irradiation is considered. Also, the failure criteria considered are plastic collapse, unstable fracture and crack penetration. For the crack growth by DHC, the failure probability was evaluated in due consideration of flaw generation time.

  2. The Application of Advancements in Computer Technology to the Control and Safety System of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chan, P. S. W. [AECL CANDU/Sheridan Park Research Community, Ontario (Canada)

    1992-04-15

    The present spatial control algorithm in CANDU reactors is based on flux synthesis from a set of parti-coloured harmonic flux modes. The design of the Rop system is also based on parti-coloured flux shapes, including both normal and abnormal reactor operating conditions. The dependency of the control and safety systems on parti-coloured data was necessitated by the slow CPU and by the scarcity of Ram which were available to the computer systems in the early seventies. Recent advancements in high speed microprocessors and high capacity Ram chips enable the development of the Pmfp computer code, which calculates reactor power distribution on-line, using diffusion theory and in-core self-powered flux detector readings as internal boundary conditions. The Pmfp based control and safety systems do not depend on parti-coloured flux shapes or preconceived reactor operating conditions.

  3. Fitness for service assessment of degraded CANDU feeder piping-Canadian regulatory expectations

    Energy Technology Data Exchange (ETDEWEB)

    Jin, John C., E-mail: john.jin@cnsc-ccsn.gc.c [Operational Engineering Assessment Div., Canadian Nuclear Safety Commission (Canada); Awad, Raoul [Operational Engineering Assessment Div., Canadian Nuclear Safety Commission (Canada)

    2011-03-15

    Allowance for the continued operation of feeder piping at some Canadian CANDU stations, which is experiencing active degradation mechanisms, has been based primarily on augmented inspection practices and conservative fitness for service assessments. The major degradation mechanisms identified to date are: pipe wall thinning due to Flow Accelerated Corrosion (FAC) and service induced cracking due to Intergranular Cracking due to Stress Corrosion Cracking (SCC) and potentially Low Temperature Creep Cracking (LTCC) mechanisms. Given that currently available industry codes and standards do not provide sufficient guidelines/criteria for assessing the degradation of feeder pipes, the Canadian Nuclear Safety Commission (CNSC) has asked the utilities to establish feeder pipe specific procedures to provide reasonable assurance that the risk associated with the feeder degradation is maintained at an acceptably low level. In response to this requirement, the Canadian CANDU industry has developed and continued to update feeder fitness for service guidelines to provide evaluation procedures and industry standard acceptance criteria for assessing the structural integrity of the feeder pipes. The scope and frequency of inspections are determined based on the results of the fitness for service assessments taking into account the relative susceptibility of feeder pipes to each specific degradation mechanism. While industry practices for the management of degraded feeder pipes have, in general, been complied with the regulatory expectations, outstanding issues still remain. Major regulatory concerns include uncertainties associated with limitations in both the inspection techniques and the mechanistic understanding of the degradation processes, which can impede inspection planning and fitness for service assessments. This paper presents the regulator's view of the current situation with respect to degradation of feeder piping, its implications for nuclear safety and the

  4. Experimental study of humidity changes on the performance of an elliptical single four-channel PEM fuel cell

    Science.gov (United States)

    Gholizadeh, Mohammad; Ghazikhani, Mohsen; Khazaee, Iman

    2017-01-01

    Humidity and humidification in a proton exchange membrane fuel cells (PEM) can significantly affect the performance of these energy generating devices. Since protons (H+) needs to be accompanied by water molecules to pass from the anode side to the cathode side, the PEM fuel cell membrane should be sufficiently wet. Low or high amount of water in the membrane can interrupt the flow of protons and thus reduce the efficiency of the fuel cell. In this context, several experimental studies and modeling have been carried out on PEM fuel cell and interesting results have been achieved. In this paper, the humidity and flow rate of gas in the anode and cathode are modified to examine its effect on fuel cell performance. The results show that the effect of humidity changing in the anode side is greater than that of the cathode so that at zero humidity of anode and 70 % humidity of the cathode, a maximum current flow of 0.512 A/cm2 for 0.12 V was obtained. However, at 70 % anode humidity and zero cathode humidity, a maximum flow of 0.86 A/cm2 for 0.13 V was obtained.

  5. Measurement of Velocity and Temperature Profiles in the 1/40 Scaled-Down CANDU-6 Moderator Tank

    Directory of Open Access Journals (Sweden)

    Hyoung Tae Kim

    2015-01-01

    Full Text Available In order to simulate the CANDU-6 moderator circulation phenomena during steady state operating and accident conditions, a scaled-down moderator test facility has been constructed at Korea Atomic Energy Institute (KAERI. In the present work an experiment using a 1/40 scaled-down moderator tank has been performed to identify the potential problems of the flow visualization and measurement in the scaled-down moderator test facility. With a transparent moderator tank model, a flow field is visualized with a particle image velocimetry (PIV technique under an isothermal state, and the temperature field is measured using a laser induced fluorescence (LIF technique. A preliminary CFD analysis is also performed to find out the flow, thermal, and heating boundary conditions with which the various flow patterns expected in the prototype CANDU-6 moderator tank can be reproduced in the experiment.

  6. Three-component U-Pu-Th fuel for plutonium irradiation in heavy water reactors

    Directory of Open Access Journals (Sweden)

    Peel Ross

    2016-01-01

    Full Text Available This paper discusses concepts for three-component fuel bundles containing plutonium, uranium and thorium for use in pressurised heavy water reactors, and cases for and against implementation of such a nuclear energy system in the United Kingdom. Heavy water reactors are used extensively in Canada, and are deploying within India and China, whilst the UK is considering the use of heavy water reactors to manage its plutonium inventory of 140 tonnes. The UK heavy water reactor proposal uses a mixed oxide (MOX fuel of plutonium in depleted uranium, within the enhanced CANDU-6 (EC-6 reactor. This work proposes an alternative heterogeneous fuel concept based on the same reactor and CANFLEX fuel bundle, with eight large-diameter fuel elements loaded with natural thorium oxide and 35 small-diameter fuel elements loaded with a MOX of plutonium and reprocessed uranium stocks from UK MAGNOX and AGR reactors. Indicative neutronic calculations suggest that such a fuel would be neutronically feasible. A similar MOX may alternatively be fabricated from reprocessed <5% enriched light water reactor fuel, such as the fuel of the AREVA EPR reactor, to consume newly produced plutonium from reprocessing, similar to the DUPIC (direct use of PWR fuel in CANDU process.

  7. 超临界水冷堆双排燃料组件子通道分析%Analysis of the sub-channel of SCWR two-row fuel assembly

    Institute of Scientific and Technical Information of China (English)

    许志红; 杨晓; 傅晟威; 杨燕华

    2012-01-01

    研究基于Cobra-IV程序,开发了适用于超临界水冷堆燃料组件分析的子通道程序.针对超临界水冷堆慢谱双排组件,进行了稳态计算,获取了相关组件热工水力参数.在此基础上,针对单一通道进行了瞬态计算,分析了燃料棒线功率变化和冷却剂流量变化条件下,超临界水冷堆燃料组件的流动和传热的动态响应,为超临界水冷堆组件的优化设计提供了参考.%Based on the COBRA-IV code, a new sub-channel code system developed for the supercritical water cooled reactor (SCWR) fuel assembly is analyzed. In order to optimize the SCWR fuel assembly design, a sub-channel analysis of two rows SCWR fuel assembly is performed, including steady-state and transient calculation. For the steady-state calculation, several channel's parameters are selected to evaluate the thermal-hydraulic performance of the fuel assemblies. Based on the steady-state results, two transient calculations (change of fuel rod power and change of coolant flow) are carried out to estimate the dynamic behavior of the fuel assemblies. The results achieved so far indicate a good applicability of the sub-channel code for the SCWR fuel assembly analysis, which is good for the future optimization of SCWR fuel assembly design.

  8. PEM燃料电池毛细压力流道拓展模型%A Capillary Extended Channel Model of PEM Fuel Cells

    Institute of Scientific and Technical Information of China (English)

    杨晓光; 叶强; 郑平

    2011-01-01

    Water flooding in porous electrodes and gas channels is the main issue hampering the performance of Proton Exchange Membrane (PEM) fuel cells. Since most traditional numerical models are based on “Mist-flow" assumption, which will greatly underestimate the liquid saturation within the cell, it is of vital importance to develop a model accounting for liquid evolution in gas channels. In light of that, we propose a capillary extended channel model and carry out a series of simulations focusing on heat and mass transport within the cell under low gas velocities. It can be drawn from the results that this model can not only overcome the setback of underestimation in liquid saturation of traditional models, but also can account for the effects of different wettability of channel walls.%多孔电极与流道内的水淹现象是制约质子交换膜(PEM)燃料电池性能提高的关键难题.传统的宏观两相模型普遍存在液态水饱和度预测值偏低的不足,而流道内气液两相雾状流的假设则是低估水淹程度的一个主要原因.本文提出了反映流道壁面浸润差异性的毛细压力流道拓展模型,并对低气速下燃料电池内的传热传质进行了模拟.计算结果表明,该模型克服了原有模型对饱和度预测值偏低的不足,不仅反映了液相水分布的贴壁效应,同时还可以考虑流道壁面差异的影响.

  9. Formation of vortex structures in channels with mass injection and their interaction with surfaces in solid-fuel rocket engines

    Science.gov (United States)

    Benderskiy, B. Ya.; Chernova, A. A.

    2015-03-01

    The topological features of the structure of combustion products flow in the flow paths with different shapes of channel cross sections at power installations are considered. The results of mathematical modeling of internal gas dynamics of the flow paths of power installations are compared with experimental data.

  10. Two-dimensional modeling of a polymer electrolyte membrane fuel cell with long flow channel. Part I. Model development

    OpenAIRE

    2015-01-01

    A two-dimensional single-phase model is developed for the steady-state and transient analysis of polymer electrolyte membrane fuel cells (PEMFC). Based on diluted and concentrated solution theories, viscous flow is introduced into a phenomenological multi-component modeling framework in the membrane. Characteristic variables related to the water uptake are discussed. A ButlereVolmer formulation of the current-overpotential relationship is developed based on an elementary mechanism of elect...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  12. Two-dimensional modeling of a polymer electrolyte membrane fuel cell with long flow channel. Part II. Physics-based electrochemical impedance analysis

    Science.gov (United States)

    Bao, Cheng; Bessler, Wolfgang G.

    2015-03-01

    The state-of-the-art electrochemical impedance spectroscopy (EIS) calculations have not yet started from fully multi-dimensional modeling. For a polymer electrolyte membrane fuel cell (PEMFC) with long flow channel, the impedance plot shows a multi-arc characteristic and some impedance arcs could merge. By using a step excitation/Fourier transform algorithm, an EIS simulation is implemented for the first time based on the full 2D PEMFC model presented in the first part of this work. All the dominant transient behaviors are able to be captured. A novel methodology called 'configuration of system dynamics', which is suitable for any electrochemical system, is then developed to resolve the physical meaning of the impedance spectra. In addition to the high-frequency arc due to charge transfer, the Nyquist plots contain additional medium/low-frequency arcs due to mass transfer in the diffusion layers and along the channel, as well as a low-frequency arc resulting from water transport in the membrane. In some case, the impedance spectra appear partly inductive due to water transport, which demonstrates the complexity of the water management of PEMFCs and the necessity of physics-based calculations.

  13. A numerical study of channel-to-channel flow cross-over through the gas diffusion layer in a PEM-fuel-cell flow system using a serpentine channel with a trapezoidal cross-sectional shape

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Lan; Oosthuizen, Patrick H. [Department of Mechanical and Materials Engineering, McLaughlin Hall, Queen' s University, Kingston, ON, K7L 3N6 (Canada); McAuley, Kim B. [Department of Chemical Engineering, Dupuis Hall, Queen' s University, Kingston, ON, K7L 3N6 (Canada)

    2006-10-15

    A numerical study of pressure distribution and flow cross-over through the gas diffusion layer (GDL) in a PEMFC flow plate using a serpentine channel system has been undertaken for the case where the channel has a trapezoidal cross-sectional shape. The flow has been assumed to be 3-D, steady, incompressible and single-phase. The flow through the porous diffusion layer has been described using the Darcy model. The governing equations have been written in dimensionless form and solved by using the commercial CFD solver, FIDAP. The results obtained indicate that: (1) the size ratio, R, of trapezoidal cross-sectional shape has a significant effect on the flow cross-over. As R increases, the flow cross-over through GDL increases; (2) the ratio R also has a significant effect on the pressure variation in the flow field for both cross-over and no cross-over cases; (3) flow cross-over has a significant influence on the pressure variation through the channel, tending to decrease the pressure drop across the channel; (4) an increase in Re number can lead to a slight increase in the flow cross-over. (author)

  14. Linear stability of a fuel channel uniformly heated considering retrofeeding by vacuum. Theoretical study; Estabilidad lineal de un canal combustible uniformemente calentado considerando retroalimentacion por vacios

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, V. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 03000 Mexico D.F. (Mexico); Salinas H, J.G.; Espinosa P, G. [Universidad Autonoma Metropolitana-Iztapalapa, 09000 Mexico D.F. (Mexico)

    2000-07-01

    The Technology, Regulation and Services Management of the National Commission of Nuclear Safety and Safeguards in coordinated form with the IPH Department of the Metropolitan Autonomous-Iztapalapa University, developed the present project to study the linear stability in a fuel channel uniformly heated with effects of retrofeeding by vacuums. In this study the methodology used in the analysis of linear stability of the nuclear reactor unit 1 at Laguna Verde power plant is described which represented by an average channel uniformly heated. The conceptual model consists of two cells which represent the two regions in which is divided the channel according to the cooling is in one and two phases, considering the boiling length dependent in the time. It is used the homogeneous flux models for describing the thermohydraulic behavior of the cooling in the two phases region. The neutron processes with the punctual model of the neutron kinetics with a group of retarded neutrons precursors are described. It is studied the behavior of the system in the frequency domain with the transfer functions obtained and it is characterized in four operation states corresponding to the four corners of the low stability zone in the map power-flow Laguna Verde power plant. For these operation states the characteristic frequency is determined and the corresponding Nyquist diagrams are obtained. The results show that the system stability depends on the power-flow relation and that the operations which implicate a reduction of this relation improve the stability of the system (reducing the power introducing control bars with constant cooling flow or increase cooling flow with bars pattern established). The obtained results with effects of retrofeeding by vacuums show that the value of the characteristic frequency is modified very little with respect to the model without retrofeeding, therefore the thermohydraulic processes seem to determine the response of the stability of the system

  15. Spent fuel characteristics & disposal considerations

    Energy Technology Data Exchange (ETDEWEB)

    Oversby, V.M.

    1996-06-01

    The fuel used in commercial nuclear power reactors is uranium, generally in the form of an oxide. The gas-cooled reactors developed in England use metallic uranium enclosed in a thin layer of Magnox. Since this fuel must be processed into a more stable form before disposal, we will not consider the characteristics of the Magnox spent fuel. The vast majority of the remaining power reactors in the world use uranium dioxide pellets in Zircaloy cladding as the fuel material. Reactors that are fueled with uranium dioxide generally use water as the moderator. If ordinary water is used, the reactors are called Light Water Reactors (LWR), while if water enriched in the deuterium isotope of hydrogen is used, the reactors are called Heavy Water reactors. The LWRs can be either pressurized reactors (PWR) or boiling water reactors (BWR). Both of these reactor types use uranium that has been enriched in the 235 isotope to about 3.5 to 4% total abundance. There may be minor differences in the details of the spent fuel characteristics for PWRs and BWRs, but for simplicity we will not consider these second-order effects. The Canadian designed reactor (CANDU) that is moderated by heavy water uses natural uranium without enrichment of the 235 isotope as the fuel. These reactors run at higher linear power density than LWRs and produce spent fuel with lower total burn-up than LWRs. Where these difference are important with respect to spent fuel management, we will discuss them. Otherwise, we will concentrate on spent fuel from LWRs.

  16. External cost assessment for nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Park, Byung Heung [Korea National University of Transportation, Chungju (Korea, Republic of); Ko, Won Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-12-15

    Nuclear power is currently the second largest power supply method in Korea and the number of nuclear power plants are planned to be increased as well. However, clear management policy for spent fuels generated from nuclear power plants has not yet been established. The back-end fuel cycle, associated with nuclear material flow after nuclear reactors is a collection of technologies designed for the spent fuel management and the spent fuel management policy is closely related with the selection of a nuclear fuel cycle. Cost is an important consideration in selection of a nuclear fuel cycle and should be determined by adding external cost to private cost. Unlike the private cost, which is a direct cost, studies on the external cost are focused on nuclear reactors and not at the nuclear fuel cycle. In this research, external cost indicators applicable to nuclear fuel cycle were derived and quantified. OT (once through), DUPIC (Direct Use of PWR SF in CANDU), PWR-MOX (PWR PUREX reprocessing), and Pyro-SFR (SFR recycling with pyroprocessing) were selected as nuclear fuel cycles which could be considered for estimating external cost in Korea. Energy supply security cost, accident risk cost, and acceptance cost were defined as external cost according to precedent and estimated after analyzing approaches which have been adopted for estimating external costs on nuclear power generation.

  17. Simulation of In-Core Dose Rates for an Offline CANDU Reactor

    Science.gov (United States)

    Gilbert, Jordan

    This thesis describes the development of a Monte Carlo simulation to predict the dose rates that will be encountered by a novel robotic inspection system for the pressure tubes of an offline CANDU reactor. Simulations were performed using the Monte Carlo N-Particle (MCNP) radiation transport code, version 6.1. The radiation fields within the reactor, even when shut down, are very high, and can cause significant damage to certain structural components and the electronics of the inspection system. Given that the robotic system will rely heavily on electronics, it is important to know the dose rates that will be encountered, in order to estimate the component lifetimes. The MCNP simulation was developed and benchmarked against information obtained from Ontario Power Generation and the Canadian Nuclear Laboratories. The benchmarking showed a good match between the simulated values and the expected values. This simulation, coupled with the accompanying user interface, represent a tool in dose field prediction that is currently unavailable. Predicted dose rates for a postulated inspection at 7 days after shutdown, with 2:5 cm of tungsten shielding around the key components, would survive for approximately 7 hours in core. This is anticipated to be enough time to perform an inspection and shows that the use of this tool can aid in designing the new inspection system.

  18. Seismic fragility analysis of a CANDU containment structure for near-fault ground motions

    Energy Technology Data Exchange (ETDEWEB)

    Choi, In Kil; Choun, Young Sun; Seo, Jeong Moon; Ahn, Seong Moon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    The R. G. 1.60 spectrum used for the seismic design of Korean nuclear power plants provides a generally conservative design basis due to its broadband nature. A survey on some of the Quaternary fault segments near Korean nuclear power plants is ongoing. It is likely that these faults will be identified as active ones. If the faults are confirmed as active ones, it will be necessary to reevaluate the seismic safety of the nuclear power plants located near these faults. The probability based scenario earthquakes were identified as near-field earthquakes. In general, the near-fault ground motion records exhibit a distinctive long period pulse like time history with very high peak velocities. These features are induced by the slip of the earthquake fault. Near-fault ground motions, which have caused much of the damage in recent major earthquakes, can be characterized by a pulse-like motion that exposes the structure to a high input energy at the beginning of the motion. It is necessary to estimate the near-fault ground motion effects on the nuclear power plant structures and components located near the faults. In this study, the seismic fragility analysis of a CANDU containment structure was performed based on the results of nonlinear dynamic time-history analyses.

  19. Nonlinear seismic behavior of a CANDU containment building subjected to near-field ground motions

    Energy Technology Data Exchange (ETDEWEB)

    Choi, In Kil; Ahn, Seong Moon; Choun, Young Sun; Seo, Jeong Moon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    The standard response spectrum proposed by US NRC has been used as a design earthquake for the design of Korean nuclear power plant structures. A survey on some of the Quaternary fault segments near Korean nuclear power plants is ongoing. It is likely that these faults will be identified as active ones. If the faults are confirmed as active ones, it will be necessary to reevaluate the seismic safety of the nuclear power plants located near the fault. Near-fault ground motions are the ground motions that occur near an earthquake fault. In general, the near-fault ground motion records exhibit a distinctive long period pulse like time history with very high peak velocities. These features are induced by the slip of the earthquake fault. Near-fault ground motions, which have caused much of the damage in recent major earthquakes, can be characterized by a pulse-like motion that exposes the structure to a high input energy at the beginning of the motion. In this study, nonlinear dynamic time-history analyses were performed to investigate the seismic behavior of a CANDU containment structure subjected to various earthquake ground motions including the near-field ground motions.

  20. Comparison Study on Thermal-Hydraulic Analysis Depending on Liquid Relief Valve Response for an Station Blackout in CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. M.; Kho, D. W. [KHNP-CRI, Daejeon (Korea, Republic of); Choi, S. H.; Moon, B. J.; Kim, S. R. [Nuclear Engineering Service and Solution Co., Daejeon (Korea, Republic of)

    2014-10-15

    The purpose of this analysis is to compare the results of thermal-hydraulic analysis depending on liquid relief valve response during a station black out (SBO) events in CANDU-6. The primary heat transport system (PHTS) behavior following the postulated SBO is analyzed using CATHENA code. In the paper, analysis was performed to evaluate the effect on coolant system where LRVs are assumed to be opened or opened according to normal open characteristics in the condition of SBO. The result showed that the primary pressure boundary is extended from LRV to DCT and the effects on primary system behavior were neglectable.

  1. Hitachi's proposed DCS solution for new build CANDU EC6 using the G-HIACS unified platform

    Energy Technology Data Exchange (ETDEWEB)

    Tan, D.; Ishii, K.; Otsuka, Y.; Uemura, K., E-mail: daisuke.tan.ye@hitachi.com [Hitachi Ltd., Infrastructure Systems Co., Ibaraki (Japan); Marko, P.E. [Hitachi Power Systems Canada Ltd., Power and Industry Div., Ontario (Canada)

    2013-07-01

    Hitachi Ltd. has developed the safe and secure functional safety DCS controller for potential new build NPP projects in the global market. Hitachi has improved the availability, maintainability, and reliability for its latest DCS systems named G-HIACS. In this latest paper on its DCS product development program, Hitachi would like to report a proposed DCS solution for new build CANDU NSP and BOP based on the G-HIACS Unified Architecture (R800FS/HSC800FS vSAFE Functional Safety Controller and R900/HSC900 General Purpose Controller) hybrid control system. (author)

  2. Assessment of System Behavior and Actions Under Loss of Electric Power For CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kang, San Ha; Moon, Bok Ja; Kim, Seoung Rae [Nuclear Engineering Service and Solution Co., Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    For the analysis, the CANDU-6 plant in Korea is considered and only the passive components are operable. The other systems are assumed to be at failed condition due to the loss of electric power. At this accident, only the inventories remained in the primary heat transport system (PHTS) and steam generator can be used for the decay heat removal. Due to the transfer of decay heat, the inventory of steam generator secondary side is discharged to the air through passive operation of main steam safety valves (MSSVs). After the steam generators are dried, the PHTS is over-pressurized and the coolant is discharged to fuelling machine vault through passive operation of degasser condenser tank relief valves (DCRVs). Under this situation, the maintenance of the integrity of PHTS is important for the protection of radionuclides release to the environment. Thus, deterministic analysis using CATHENA code is carried out for the simulation of the accident and the appropriate operator action is considered. The loss of electric power results in the depletion of steam generator inventory which is necessary for the decay heat removal. If only the passive system is credited, the PT can be failed after the steam generator is depleted. For the prevention of the PT failure, the feedwater should be supplied to the steam generator before 4,800s after the accident. The feedwater can be supplied using water in dousing tank if the steam generators are depressurized. The decay heat from the core is removed through natural circulation if the feedwater can be supplied continuously.

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jun Su; Jeong, Seung Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

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

  4. Performance enhancement of polymer electrolyte fuel cells by combining liquid removal mechanisms of a gas diffusion layer with wettability distribution and a gas channel with microgrooves

    Science.gov (United States)

    Utaka, Yoshio; Koresawa, Ryo

    2016-08-01

    Although polymer electrolyte fuel cells (PEFCs) are commercially available, there are still many problems that need to be addressed to improve their performance and increase their usage. At a high current density, generated water accumulates in the gas diffusion layer and in the gas channels of the cathode. This excess water obstructs oxygen transport, and as a result, cell performance is greatly reduced. To improve the cell performance, the effective removal of the generated water and the promotion of oxygen diffusion in the gas diffusion layer (GDL) are necessary. In this study, two functions proposed in previous reports were combined and applied to a PEFC: a hybrid GDL to form an oxygen diffusion path using a wettability distribution and a gas separator with microgrooves to enhance liquid removal. For a PEFC with a hybrid GDL and a gas separator with microgrooves, the concentration overvoltage of the PEFC was reduced, and the current density limit and maximum power density were increased compared with a conventional PEFC. Moreover, the stability of the cell voltage was markedly improved.

  5. Review of design criteria and safety analysis of safety class electric building for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. HANARO fuel test loop was designed for CANDU and PWR fuel testing. Safety related system of Fuel Test Loop such as emergency cooling water system, component cooling water system, safety ventilation system, high energy line break mitigation system and remote control room was required 1E class electric supply to meet the safety operation in accordance with related code. Therefore, FTL electric building was designed to construction and install the related equipment based on seismic category I. The objective of this study is to review the design criteria and analysis the safety function of safety class electric building for fuel test loop, and this results will become guidance for the irradiation testing in future. (author). 10 refs., 6 tabs., 30 figs.

  6. Enhancement of proton exchange membrane fuel cell performance using a novel tapered gas channel%以新型渐缩式流道增进质子交换膜燃料电池性能

    Institute of Scientific and Technical Information of China (English)

    钟振忠; 陈俊勋; 庄平吉

    2009-01-01

    Based on use of multi-dimensional models,this investigation simulates the performance of a proton exchange membranc fuel eell by varying the channel pattern.In the one-dimensionaI model,the porosity of the gas diffusion Iayer is 0.3.The modcl reveals the water vapor distribution of the fucI cell and demonstrates that the amount of water vapor increases linearly with the reduction reaction adjacent to the gas channeI and the gas diflusion layer.Secondly,four hovel tapered gas channels are simulated by a two.dimensionaI model.The modeI considers the distributions of oxygen.the pressure drop,the amount of water vapor distribution and the polarization curves.The results indicate that as the channel depth decreases.the oxygen in the tapered gas channel can be acceleratcd and forced into the gas diflusion layer to improve the cell performance.The three-dimcnsional model is employed to simulate the phcnomcnon associated with four novel tapered gas channels.The results also show that the best performancc is realized in the Ieast tapered gas channel.Finally.an experimentally determined mechanism is found to be consistent with the results of the simulation.

  7. Feasibility investigations on a novel micro-manufacturing process for fabrication of fuel cell bipolar plates: Internal pressure-assisted embossing of micro-channels with in-die mechanical bonding

    Energy Technology Data Exchange (ETDEWEB)

    Koc, Muammer [NSF I/UCR Center for Precision Forming (CPF), Department of Mechanical Engineering, Virginia Commonwealth University (VCU), Richmond, VA (United States); Mahabunphachai, Sasawat [NSF I/UCR Center for Precision Forming (CPF), Department of Mechanical Engineering, Virginia Commonwealth University (VCU), Richmond, VA (United States); Department of Mechanical Engineering, University of Michigan, Ann Arbor, MI (United States)

    2007-10-25

    In this paper, we present the results of our studies on conceptual design and feasibility experiments towards development of a novel hybrid manufacturing process to fabricate fuel cell bipolar plates that consists of multi-array micro-channels on a large surface area. The premises of this hybrid micro-manufacturing process stem from the use of an internal pressure-assisted embossing process (cold or warm) combined with mechanical bonding of double bipolar plates in a single-die and single-step operation. Such combined use of hydraulic and mechanical forming forces and in-process bonding will (a) enable integrated forming of micro-channels on both surfaces (as anode and cathode flow fields) and at the middle (as cooling channels), (b) reduce the process steps, (c) reduce variation in dimensional tolerances and surface finish, (d) increase the product quality, (e) increase the performance of fuel cell by optimizing flow-field designs and ensuring consistent contact resistance, and (f) reduce the overall stack cost. This paper explains two experimental investigations that were performed to characterize and evaluate the feasibility of the conceptualized manufacturing process. The first investigation involved hydroforming of micro-channels using thin sheet metals of SS304 with a thickness of 51 {mu}m. The width of the channels ranged from 0.46 to 1.33 mm and the height range was between 0.15 and 0.98 mm. Our feasibility experiments resulted in that different aspect ratios of micro-channels could be fabricated using internal pressure in a controllable manner although there is a limit to very sharp channel shapes (i.e., high aspect ratios with narrow channels). The second investigation was on the feasibility of mechanical bonding of thin sheet metal blanks. The effects of different process and material variables on the bond quality were studied. Successful bonding of various metal blanks (Ni201, Al3003, and SS304) was obtained. The experimental results from both

  8. Thermal Analysis of CANDU6-Moderator System for Loss of Cooling%CANDU6慢化剂系统丧失冷却情况下的温度分析

    Institute of Scientific and Technical Information of China (English)

    徐珍

    2012-01-01

    The coolant system and moderator system of CANDU6 are independent. The prompt neutrons are moderated as thermal neutrons by the moderator and the continuous nuclear fission in the reactor is maintained. At the same time the moderator system supplies the heat sink for the heat produced by the neutrons moderation. During the in-service maintenance of plant, the standby RCW which will only cool down reactor coolant system operates instead of RCW and can not supply heat sink for moderator system heat exchanger. As the result, the moderator system will lose heat sink during the operation of standby RCW. To estimate the moderator temperature, was compared with the experiment data and evaluated in this paper. the thermal analysis of moderator system for loss of cooling the system failure caused by the temperature raising wasevaluated in this paper.%压力管卧式重水反应堆(CANDU6)具有相互独立的冷却剂系统和慢化剂系统。慢化剂系统将堆芯高能裂变中子慢化到能维持持续裂变所需的热中子水平,并将慢化中子过程中产生的热量带出。在反应堆大修期间,需要对再循环冷却水系统(RCW)进行检修,则需要并投入其备用系统,但是RCW备用系统仅对反应堆冷却剂系统进行冷却,不提供慢化剂系统热交换器冷却水。所以在RCW备用系统投入的情况下,慢化剂系统丧失冷却。为判断在此情况下慢化剂的温度变化情况,本文对CANDU6大修期间慢化剂系统丧失冷却情况下的温度变化进行分析并与试验结果进行比较,评估是否会由于温度过高而导致系统失效。

  9. Optimum design of the slotted-interdigitated channels flow field for proton exchange membrane fuel cells with consideration of the gas diffusion layer intrusion

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Linfa; Mai, Jianming; Hu, Peng; Lai, Xinmin; Lin, Zhongqin [State Key Laboratory of Mechanical System and Vibration, Shanghai Jiao Tong University, Shanghai 200240 (China)

    2011-05-15

    The design of the flow field greatly affects the flow distribution and the final performance of the proton exchange membrane fuel cell (PEMFC) system. The clamping force between the gas diffusion layer (GDL) and the flow field plate (FFP) will cause the inhomogeneous compression of the GDL. Then the GDL will be intruded into the reactant gas channels and eventually change the flow distribution. This paper presents a study on the effect of the intrusion of the GDL on the flow field in a PEMFC, and tries to explain the reason for poor performance of the previous design of the flow field other than those have been studied in other papers such as GDL roughness, porosity, etc. First of all, a linear analytical model is used to analyze the sensitivities of the flow field to the GDL intrusion, and then used to estimate the effect of the GDL intrusion on the flow field distribution. Secondly, a multi-objective optimization model is proposed to eliminate the nonuniform distribution in the flow field with the GDL intrusion taken into consideration. Subsequently, three different designs are analyzed and compared with each other as a demonstration to show the effect of the GDL intrusion. From the analysis results, it is recommended that the effect of the GDL intrusion on the multi-depth flow field should be taken into consideration and the flow field should be insensitive to the GDL intrusion to obtain high robust performance. The results obtained in the study provide the designer some useful guidelines in the concept design of flow field configurations. (author)

  10. A study on the environmental friendliness of nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. J.; Lee, B. H.; Lee, S. Y.; Lim, C. Y.; Choi, Y. S.; Lee, Y. E.; Hong, D. S.; Cheong, J. H; Park, J. B.; Kim, K. K.; Cheong, H. Y; Song, M. C; Lee, H. J. [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of)

    1998-01-01

    The purpose of this study is to develop methodologies for quantifying environmental and socio-political factors involved with nuclear fuel cycle and finally to evaluate nuclear fuel cycle options with special emphasis given to the factors. Moreover, methodologies for developing practical radiological health risk assessment code system will be developed by which the assessment could be achieved for the recycling and reuse of scrap materials containing residual radioactive contamination. Selected scenarios are direct disposal, DUPIC(Direct use of PWR spent fuel in CANDU), and MOX recycle, land use, radiological effect, and non-radiological effect were chosen for environmental criteria and public acceptance and non-proliferation of nuclear material for socio-political ones. As a result of this study, potential scenarios to be chosen in Korea were selected and methodologies were developed to quantify the environmental and socio-political criteria. 24 refs., 27 tabs., 29 figs. (author)

  11. Remote fabrication and irradiation test of recycled nuclear fuel prepared by the oxidation and reduction of spent oxide fuel

    Science.gov (United States)

    Jin Ryu, Ho; Chan Song, Kee; Il Park, Geun; Won Lee, Jung; Seung Yang, Myung

    2005-02-01

    A direct dry recycling process was developed in order to reuse spent pressurized light water reactor (LWR) nuclear fuel in CANDU reactors without the separation of sensitive nuclear materials such as plutonium. The benefits of the dry recycling process are the saving of uranium resources and the reduction of spent fuel accumulation as well as a higher proliferation resistance. In the process of direct dry recycling, fuel pellets separated from spent LWR fuel rods are oxidized from UO2 to U3O8 at 500 °C in an air atmosphere and reduced into UO2 at 700 °C in a hydrogen atmosphere, which is called OREOX (oxidation and reduction of oxide fuel). The pellets are pulverized during the oxidation and reduction processes due to the phase transformation between cubic UO2 and orthorhombic U3O8. Using the oxide powder prepared from the OREOX process, the compaction and sintering processes are performed in a remote manner in a shielded hot cell due to the high radioactivity of the spent fuel. Most of the fission gas and volatile fission products are removed during the OREOX and sintering processes. The mini-elements fabricated by the direct dry recycling process are irradiated in the HANARO research reactor for the performance evaluation of the recycled fuel pellets. Post-irradiation examination of the irradiated fuel showed that microstructural evolution and fission gas release behavior of the dry-recycled fuel were similar to high burnup UO2 fuel.

  12. BEAM 1.7: development for modelling fuel element and bundle buckling strength

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, G.; Xu, S.; Xu, Z.; Paul, U.K. [Atomic Energy of Canada, Mississauga, Ontario (Canada)

    2010-07-01

    This paper describes BEAM, an AECL developed computer program, used to assess mechanical integrity of CANDU fuel bundles. The BEAM code has been developed to satisfy the need for buckling strength analysis of fuel bundles. Buckling refers to the phenomenon where a compressive axial load is large enough that a small lateral load can cause large lateral deflections. The buckling strength refers to the critical compressive axial load at which lateral instability is reached. The buckling strength analysis has practical significance for the design of fuel bundles, where the buckling strength of a fuel element/bundle is assessed so that the conditions leading to bundle jamming in the pressure tube are excluded. This paper presents the development and qualification of the BEAM code, with emphasis on the theoretical background and code implementation of the newly developed fuel element/bundle buckling strength model. (author)

  13. Modelling material effects on flow-accelerated corrosion in primary CANDU coolant and secondary reactor feed-water

    Energy Technology Data Exchange (ETDEWEB)

    Phromwong, P.; Lister, D., E-mail: c7r13@unb.ca [Univ. of New Brunswick, Dept. of Chemical Engineering, Fredericton, New Brunswick (Canada); Uchida, S. [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan)

    2012-07-01

    The effects of chromium content on flow-accelerated corrosion (FAC) of carbon steel have been predicted very well by including a passivating layer, which is a chromium-dependent diffusion barrier at the metal-oxide interface. By adjusting the properties of the chromium-dependent layer, described with a Passivation Parameter (PP), we can predict the FAC of carbon steel of different chromium contents in typical reactor feed-water environments (140{sup o}C and neutral or ammoniated chemistry). The model and an appropriate PP are also applied to the environment typical of carbon-steel feeders in the primary coolant of a CANDU reactor (310{sup o}C and lithiated chemistry). The model predicts FAC rate very well (with a deviation of 10% or less) in both situations. (author)

  14. Probabilistic seismic safety assessment of a CANDU 6 nuclear power plant including ambient vibration tests: Case study

    Energy Technology Data Exchange (ETDEWEB)

    Nour, Ali [Hydro Québec, Montréal, Québec H2L4P5 (Canada); École Polytechnique de Montréal, Montréal, Québec H3C3A7 (Canada); Cherfaoui, Abdelhalim; Gocevski, Vladimir [Hydro Québec, Montréal, Québec H2L4P5 (Canada); Léger, Pierre [École Polytechnique de Montréal, Montréal, Québec H3C3A7 (Canada)

    2016-08-01

    Highlights: • In this case study, the seismic PSA methodology adopted for a CANDU 6 is presented. • Ambient vibrations testing to calibrate a 3D FEM and to reduce uncertainties is performed. • Procedure for the development of FRS for the RB considering wave incoherency effect is proposed. • Seismic fragility analysis for the RB is presented. - Abstract: Following the 2011 Fukushima Daiichi nuclear accident in Japan there is a worldwide interest in reducing uncertainties in seismic safety assessment of existing nuclear power plant (NPP). Within the scope of a Canadian refurbishment project of a CANDU 6 (NPP) put in service in 1983, structures and equipment must sustain a new seismic demand characterised by the uniform hazard spectrum (UHS) obtained from a site specific study defined for a return period of 1/10,000 years. This UHS exhibits larger spectral ordinates in the high-frequency range than those used in design. To reduce modeling uncertainties as part of a seismic probabilistic safety assessment (PSA), Hydro-Québec developed a procedure using ambient vibrations testing to calibrate a detailed 3D finite element model (FEM) of the containment and reactor building (RB). This calibrated FE model is then used for generating floor response spectra (FRS) based on ground motion time histories compatible with the UHS. Seismic fragility analyses of the reactor building (RB) and structural components are also performed in the context of a case study. Because the RB is founded on a large circular raft, it is possible to consider the effect of the seismic wave incoherency to filter out the high-frequency content, mainly above 10 Hz, using the incoherency transfer function (ITF) method. This allows reducing significantly the non-necessary conservatism in resulting FRS, an important issue for an existing NPP. The proposed case study, and related methodology using ambient vibration testing, is particularly useful to engineers involved in seismic re-evaluation of

  15. Hybrid Monte Carlo deterministic and probabilistic core assessment for flaws and leak-before break for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shi, P.; Mok, D.H.B. [AMEC NSS, Toronto, Ontario (Canada)

    2011-07-01

    Even though pressure tubes are major components of a CANDU reactor, only small proportions of pressure tubes are sampled for in-service inspections due to execution cost, outage duration, and site cumulative radiation exposure limits. In general, a realistic core assessment was not carried out based on all known information related to in-service degradation mechanisms. Recently, a hybrid deterministic and probabilistic core assessment (HDPCA) has been introduced to address the uncertainties associated with uninspected pressure tubes and diverse degradation mechanisms. In the present paper, the HDPCA was carried out for a CANDU unit based on cumulative operating experience and history in order to satisfy the requirements of Clause 7 of CSA Standard N285.8 by considering the uncertainties associated with the estimated distribution parameters, the limited inspected data, and pressure tube properties. The HDPCA is composed of two parts: a simulation part and a deterministic evaluation part. The outcome of the core assessment is the expected pressure tube failure frequency due to pressure tube flaws. In the simulations, pressure tube material properties were sampled from distributions derived from material surveillance and testing programs. The flaw dimensions and intensities were sampled from distributions fitted to in-service inspection data. The pressure tubes were then populated with flaws. Each simulated flaw was evaluated for DHC initiation under constant loading conditions. When Delayed Hydride Cracking initiation from a flaw was predicted, the pressure tube was evaluated for rupture in the Leak-Before-Break evaluation. Based on all the predicted pressure tube ruptures from simulations, the failure frequency was calculated on an annual basis. The largest expected mean and the 95% upper bound of the mean failure frequencies for any evaluation subinterval to the end of pressure tube design life of 210,000 EFPH are significantly below the allowable failure frequency

  16. Severe accident analysis of a station blackout accident using MAAP-CANDU for the Point Lepreau station refurbishment project level 2 PSA

    Energy Technology Data Exchange (ETDEWEB)

    Brown, M.J.; Petoukhov, S.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station, using the MAAP-CANDU code to simulate the progression of severe core damage accidents and fission product releases. Five representative severe accidents were selected: Station Blackout, Small Loss-of-Coolant, Stagnation Feeder Break, Steam Generator Tube Rupture, and Shutdown State. Analysis results for the reference station blackout accident are discussed in this paper. (author)

  17. Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography

    CERN Document Server

    Jonkmans, G; Jewett, C; Thompson, M

    2012-01-01

    This paper explores the use of cosmic ray muons to image the contents of shielded containers and detect high-Z special nuclear materials inside them. Cosmic ray muons are a naturally occurring form of radiation, are highly penetrating and exhibit large scattering angles on high Z materials. Specifically, we investigated how radiographic and tomographic techniques can be effective for non-invasive nuclear waste characterization and for nuclear material accountancy of spent fuel inside dry storage containers. We show that the tracking of individual muons, as they enter and exit a structure, can potentially improve the accuracy and availability of data on nuclear waste and the contents of Dry Storage Containers (DSC) used for spent fuel storage at CANDU plants. This could be achieved in near real time, with the potential for unattended and remotely monitored operations. We show that the expected sensitivity, in the case of the DSC, exceeds the IAEA detection target for nuclear material accountancy.

  18. Evolution of the MACSTOR{trademark} dry spent fuel storage system

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F.E.; Joubert, W.M. [Atomic Energy of Canada Ltd., Montreal, Quebec (Canada)

    1995-12-31

    The MACSTOR{trademark} (Modular Air-Cooled Canister Storage) system was developed by Atomic Energy of Canada Limited (AECL) for the interim storage of spent fuel discharged by light water reactors. It is a hybrid system which combines the operational economies of metal cask technology with the capital economies of concrete technology. The system includes all the necessary equipment to transfer spent fuel from a storage pool to an independent interim dry spent fuel storage site. After presenting a description of the system and a brief history of its development, the paper addresses its thermal performance and modeling for various design configurations. Finally, a brief summary of the experience being gained during the implementation of a MACSTOR{trademark} system modified for CANDU spent fuel at the Gentilly-2 NPP in Quebec is presented. It includes progress made in licensing activities and in public hearings pertinent to the initiation of the project.

  19. On the nuclear oxide fuel densification, swelling and thermal re-sintering

    Energy Technology Data Exchange (ETDEWEB)

    Paraschiv, M.C. E-mail: mariusparaschiv@mail.rtns.ro; Paraschiv, A.; Grecu, V.V

    2002-04-01

    A method of calculating the fuel densification and swelling as a result of only the initial fuel porosity has been developed. Well-characterized UO{sub 2} fuel pellets, from the Romanian CANDU type fuel fabrication route, were used to fit the specific pore volume distribution resulted from the sintering process to a lognormal distribution function and thermal re-sintering tests up to 280 h at 1700 deg. C were done to fit the coefficients of vacancies diffusion and the self-diffusion of uranium, required by the model. Careful analyses proved that the irradiation-induced resolution is strongly dependent of the hydrostatic pressure and a corrected formula has been proposed. Analyses of the time evolution of the pore size distribution during re-sintering tests and annealing tests similar with some as described in the open literature were also done.

  20. Fuel enrichment and temperature distribution in nuclear fuel rod in (D-T) driven hybrid reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Osman, Ypek [Suleyman Demirel Universitesi Muhendislik-Mimarlyk Fakultesi, Isparta (Turkey)

    2001-07-01

    In this study, melting point of the fuel rod and temperature distribution in nuclear fuel rod are investigated for different coolants under various first wall loads (P{sub w}, =5, 6, 7, 8, 9, and 10 MWm{sup -2}) in Fusion-Fission reactor fueled with 50%LWR +50%CANDU. The fusion source of neutrons of 14.1 MeV is simulated by a movable target along the main axis of cylindrical geometry as a line source. In addition, the fusion chamber was thought as a cylindrical cavity with a diameter of 300 cm that is comparatively small value. The fissile fuel zone is considered to be cooled with four different coolants, gas, flibe (Li{sub 2}BeF{sub 4}), natural lithium (Li), and eutectic lithium (Li{sub 17}Pb{sub 83}). Investigations are observed during 4 years for discrete time intervals of{delta}t= 0.5 month and by a plant factor (PF) of 75%. Volumetric ratio of coolant-to fuel is 1:1, 45.515% coolant, 45.515% fuel, 8.971% clad, in fuel zone. (author)

  1. Analytical solutions for the temperature field in a 2D incompressible inviscid flow through a channel with walls of solid fuel

    Directory of Open Access Journals (Sweden)

    Sorin BERBENTE

    2011-12-01

    Full Text Available A gas (oxidizer flows between two parallel walls of solid fuel. A combustion is initiated: the solid fuel is vaporized and a diffusive flame occurs. The hot combustion products are submitted both to thermal diffusion and convection. Analytical solutions can be obtained both for the velocity and temperature distributions by considering an equivalent mean temperature where the density and the thermal conductivity are evaluated. The main effects of heat transfer are due to heat convection at the flame. Because the detailed mechanism of the diffusion flame is not introduced the reference chemical reaction is the combustion of premixed fuel with oxidizer in excess. In exchange the analytical solution is used to define an ideal quasi-uniform combustion that could be realized by an n adequate control. The given analytical closed solutions prove themselves flexible enough to adjust the main data of some existing experiments and to suggest new approaches to the problem.

  2. Development of finite element models for the study of ageing effects in CANDU 6 concrete containment buildings

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Y.; Jaffer, S., E-mail: Yuqing.Ding@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    In nuclear power plants (NPPs), concrete containment buildings (CCBs) provide the final physical barrier against the release of radioactive materials into the environment and protect the nuclear structures housed within the containment building. CCBs have to be maintained to ensure leak tightness and sound structural integrity for the safe operation throughout the life of NPPs. However, the integrity of CCBs may be affected by the ageing of its concrete, post-tensioning cables and reinforcing bars (rebars). Finite element models (FEMs) of CANDU 6 CCBs have been developed using 2 independent finite element programs for the study of the effect of ageing of CCBs. These FEMs have been validated using multiple-source data and have been used for preliminary analyses of the effect of thermal load and ageing degradation on the concrete structure. The modelling assumptions and simplifications, approach, and validation are discussed in this paper. The preliminary analyses for temperature effects and potential applications to the study of ageing degradation in CCBs using the FEMs are briefly introduced. (author)

  3. CANDU堆应用RU的PWR/CANDU联合核燃料循环的研究%Study of RU Utilization in CANDU Reactor-an Advanced Nuclear Fuel Cycle of PWR/CANDU Synergism

    Institute of Scientific and Technical Information of China (English)

    霍小东; 谢仲生

    2003-01-01

    对压水堆乏燃料后处理回收铀(RU)在秦山三期CANDU堆中应用的可行性和经济性进行分析.使用ORIGEN2程序,对后处理回收铀在生产后放置不同时间后核素的成份和放射性活度进行了计算.证明RU燃料元件生产的放射性水平是可以接受的.使用DRAGON/DONJON程序对应用RU的秦山三期CANDU堆的时均堆芯和瞬时堆芯校验分析表明:采用简单的2燃耗区,2、4棒束的换料方案能满足最大通道功率、最大棒束功率限制.通过放射性分析和堆芯物理分析可以看出,秦山三期CANDU堆在不改变堆芯结构及运行模式的条件下,从天然铀(NU)燃料过渡到RU燃料是可行的.通过对秦山三期CANDU堆应用RU的经济性分析,可以看出PWR/CANDU联合核燃料循环的策略既可节约铀资源(23%),提高燃料的能量输出(41%),又减少了废燃料的处置量(66%),可大大降低核电成本.

  4. A secondary fuel removal process: plasma processing

    Energy Technology Data Exchange (ETDEWEB)

    Min, J. Y.; Kim, Y. S. [Hanyang Univ., Seoul (Korea, Republic of); Bae, K. K.; Yang, M. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-07-01

    Plasma etching process of UO{sub 2} by using fluorine containing gas plasma is studied as a secondary fuel removal process for DUPIC (Direct Use of PWR spent fuel Into Candu) process which is taken into consideration for potential future fuel cycle in Korea. CF{sub 4}/O{sub 2} gas mixture is chosen for reactant gas and the etching rates of UO{sub 2} by the gas plasma are investigated as functions of CF{sub 4}/O{sub 2} ratio, plasma power, substrate temperature, and plasma gas pressure. It is found that the optimum CF{sub 4}/O{sub 2} ratio is around 4:1 at all temperatures up to 400 deg C and the etching rate increases with increasing r.f. power and substrate temperature. Under 150W r.f. power the etching rate reaches 1100 monolayers/min at 400 deg C, which is equivalent to about 0.5mm/min. (author).

  5. Development of equipment for fabricating DUPIC fuel powder

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki Ho; Yang, M. S.; Park, J. J.; Lee, J. W.; Kim, J. H.; Cho, K. H.; Lee, D. Y.; Lee, Y. S.; Na, S. H

    1999-06-01

    The powder fabrication processes, as the first stage of manufacturing DUPIC (Direct Use of PWR spent fuel In CANDU) fuel, consist of the slitting of spent PWR fuel rods, REOX (Oxidation and REduction of Oxide Fuels) processing to produce the powder feedstock, the milling of the produced powder, the granulation of the milled powder, and the mixing of the granulated powder with pressing lubricants. All these processes should be conducted by remote means in a hot-cell environment where the direct human access is limited to the strictest minimum due to the high radioactivity. This report describe the development of the equipment for fabricating DUPIC fuel powder. These equipment are Slitting Machine, Oxidation and Reduction (OREOX) Furnace, Mill, Roll Compactor, and Mixer. Remote design concept was applied to all the equipment for use in the M6 hot-cell of the IMEF. Mechanical design considerations and capabilities of the equipment for remote operation and maintenance are presented. First prototypes were developed and installed in the DUPIC full scale mock-up and tested using a master-slave manipulator. Redesign and reconstruction were made on each equipment based on mock-up test results. The remote technology acquired through this research was utilized in developing other equipment for DUPIC fuel fabrication, thereby improving safety and increasing productivity. This technology could also be extended to the area of remote handling equipment development for use in hazardous environments. (author). 14 refs., 9 tabs., 21 figs.

  6. Nonproliferation and safeguards aspects of the DUPIC fuel cycle concept

    Energy Technology Data Exchange (ETDEWEB)

    Persiani, P. K. [Argonne National Lab., IL (United States)

    1997-07-01

    The purpose of the study is to comment on the proliferation characteristic profiles of some of the proposed fuel cycle alternatives to help ensure that nonproliferation concerns are introduced into the early stages of a fuel cycle concept development program, and to perhaps aid in the more effective implementation of the international nonproliferation regime initiative and safeguards systems. Alternative recycle concepts proposed by several countries involve the recycle of spent fuel without the separation of plutonium from uranium and fission products. The concepts are alternatives to either the direct long-term storage deposition of or the purex reprocessing of the spent fuels. The alternate fuel cycle concepts reviewed include: the dry-recycle processes such as the direct use of reconfigured PWR spent fuel assemblies into CANDU reactors(DUPIC); low-decontamination, single-cycle co-extraction of fast reactor fuels in a wet-purex type of reprocessing; and on a limited scale the thorium-uranium fuel cycle. The nonproliferation advantages usually associated with the above non-separation processes are: the highly radioactive spent fuel presents a barrier to the physical diversion of the nuclear material; avoid the need to dissolve and chemically separate the plutonium from the uranium and fission products; and that the spent fuel isotopic quality of the plutonium vector is further degraded. Although the radiation levels and the need for reprocessing may be perceived as barriers to the terrorist or the subnational level of safeguards, the international level of nonproliferation concerns is addressed primarily by material accountancy and verification activities. On the international level of nonproliferation concerns, the non-separation fuel cycle concepts involved have to be evaluated on the bases of the impact the processes may have on nuclear materials accountancy. (author).

  7. CHF Enhancement of Advanced 37-Element Fuel Bundles

    Directory of Open Access Journals (Sweden)

    Joo Hwan Park

    2015-01-01

    Full Text Available A standard 37-element fuel bundle (37S fuel bundle has been used in commercial CANDU reactors for over 40 years as a reference fuel bundle. Most CHF of a 37S fuel bundle have occurred at the elements arranged in the inner pitch circle for high flows and at the elements arranged in the outer pitch circle for low flows. It should be noted that a 37S fuel bundle has a relatively small flow area and high flow resistance at the peripheral subchannels of its center element compared to the other subchannels. The configuration of a fuel bundle is one of the important factors affecting the local CHF occurrence. Considering the CHF characteristics of a 37S fuel bundle in terms of CHF enhancement, there can be two approaches to enlarge the flow areas of the peripheral subchannels of a center element in order to enhance CHF of a 37S fuel bundle. To increase the center subchannel areas, one approach is the reduction of the diameter of a center element, and the other is an increase of the inner pitch circle. The former can increase the total flow area of a fuel bundle and redistributes the power density of all fuel elements as well as the CHF. On the other hand, the latter can reduce the gap between the elements located in the middle and inner pitch circles owing to the increasing inner pitch circle. This can also affect the enthalpy redistribution of the fuel bundle and finally enhance CHF or dry-out power. In this study, the above two approaches, which are proposed to enlarge the flow areas of the center subchannels, were considered to investigate the impact of the flow area changes of the center subchannels on the CHF enhancement as well as the thermal characteristics by applying a subchannel analysis method.

  8. Semi-annual status report of the Canadian Nuclear Fuel Waste Management Program, April 1--September 30, 1991

    Energy Technology Data Exchange (ETDEWEB)

    Wright, E.D. [comp.

    1992-02-01

    This report is the eleventh in a series of semi-annual status reports on the research and development program for the safe management and disposal of Canada's nuclear fuel waste. it describes progress achieved in the three major subprograms, engineered systems, natural systems and performance assessment, from 1991 April 1 to September 30. It also gives a brief description of the activities being carried out in preparation for the public and governmental review of the disposal concept. Since 1987, this program has been jointly funded by AECL and Ontario Hydro under the auspices of the CANDU Owners Group (COG).

  9. Severe accident analysis of a small LOCA accident using MAAP-CANDU support level 2 PSA for the Point Lepreau station refurbishment project

    Energy Technology Data Exchange (ETDEWEB)

    Petoukhov, S.M.; Brown, M.J.; Mathew, P.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station. The MAAP4-CANDU code was used to calculate the progression of postulated severe core damage accidents and fission product releases. Five representative severe core damage accidents were selected: Station Blackout, Small Loss-of-Coolant Accident, Stagnation Feeder Break, Steam Generator Tube Rupture, and Shutdown State Accident. Analysis results for only the reference Small LOCA Accident scenario (which is a very low probability event) are discussed in this paper. (author)

  10. Radionuclide Release after LBLOCA with Loss of Class IV Power Accident in CANDU-6 Plant

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hoon [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    A large break in a pipe train of a primary heat transport system discharges coolant, which has high energy and large mass, into the containment building. Reactor shutdown and emergency core cooling water will limit the fuel cladding failure, but cannot prevent it entirely. The containment building is the last barrier of radionuclide release to the environment. Containment isolation and pressure suppression by dousing and local air cooler reduce the amount of radionuclide release to the environment. The objective of containment behavior analysis for large break loss of coolant with loss of class IV power accident is to assess the amount of radionuclide release to the ambient atmosphere. Radionuclide release rates in this event, with all safety system available, that is, the containment building is intact, as well as with containment system impairment, are analyzed with GOTHIC and SMART code

  11. ACR fuel storage analysis: finite element heat transfer analysis of dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Khair, K.; Baset, S.; Millard, J. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2006-07-01

    Over the past decade Atomic Energy of Canada Limited (AECL) has designed and licensed air-cooled concrete structures used as above ground dry storage containers (MACSTOR) to store irradiated nuclear fuel from CANDU plants. A typical MACSTOR 200 module is designed to store 12,000 bundles in 20 storage cylinders. MACSTOR 200 modules are in operation at Gentilly-2 in Canada and at Cernavoda in Romania. The MACSTOR module is cooled passively by natural convection and by conduction through the concrete walls and roof. Currently AECL is designing the Advanced Candu Reactor (ACR) with CANFLEX slightly enriched uranium fuel to be used. AECL has initiated a study to explore the possibility of storing the irradiated nuclear fuel from ACR in MACSTOR modules. This included work to consider ways of minimizing footprint both in the spent fuel storage bay and in the dry storage area. The commercial finite element code ANSYS has been used in this study. The FE model is used to complete simulations with the higher heat source using the same concrete structural dimensions to assess the feasibility of using the MACSTOR design for storing the ACR irradiated fuel. This paper presents the results of the analysis. The results are used to confirm the possibility of using, with minimal changes to the design of the storage baskets and the structure, the proven design of the MACSTOR 200 containment to store the ACR fuel bundles with higher enrichment and burnup. This has thus allowed us to confirm conceptual feasibility and move on to investigation of optimization. (author)

  12. Dynamic imaging of free cytosolic ATP concentration during fuel sensing by rat hypothalamic neurones: evidence for ATP-independent control of ATP-sensitive K(+) channels.

    Science.gov (United States)

    Ainscow, Edward K; Mirshamsi, Shirin; Tang, Teresa; Ashford, Michael L J; Rutter, Guy A

    2002-10-15

    Glucose-responsive (GR) neurons from hypothalamic nuclei are implicated in the regulation of feeding and satiety. To determine the role of intracellular ATP in the closure of ATP-sensitive K(+) (K(ATP)) channels in these cells and associated glia, the cytosolic ATP concentration ([ATP](c)) was monitored in vivo using adenoviral-driven expression of recombinant targeted luciferases and bioluminescence imaging. Arguing against a role for ATP in the closure of K(ATP) channels in GR neurons, glucose (3 or 15 mM) caused no detectable increase in [ATP](c), monitored with cytosolic luciferase, and only a small decrease in the concentration of ATP immediately beneath the plasma membrane, monitored with a SNAP25-luciferase fusion protein. In contrast to hypothalamic neurons, hypothalamic glia responded to glucose (3 and 15 mM) with a significant increase in [ATP](c). Both neurons and glia from the cerebellum, a glucose-unresponsive region of the brain, responded robustly to 3 or 15 mM glucose with increases in [ATP](c). Further implicating an ATP-independent mechanism of K(ATP) channel closure in hypothalamic neurons, removal of extracellular glucose (10 mM) suppressed the electrical activity of GR neurons in the presence of a fixed, high concentration (3 mM) of intracellular ATP. Neurons from both brain regions responded to 5 mM lactate (but not pyruvate) with an oligomycin-sensitive increase in [ATP](c). High levels of the plasma membrane lactate-monocarboxylate transporter, MCT1, were found in both cell types, and exogenous lactate efficiently closed K(ATP) channels in GR neurons. These data suggest that (1) ATP-independent intracellular signalling mechanisms lead to the stimulation of hypothalamic neurons by glucose, and (2) these effects may be potentiated in vivo by the release of lactate from neighbouring glial cells.

  13. Dynamic imaging of free cytosolic ATP concentration during fuel sensing by rat hypothalamic neurones: evidence for ATP-independent control of ATP-sensitive K+ channels

    Science.gov (United States)

    Ainscow, Edward K; Mirshamsi, Shirin; Tang, Teresa; Ashford, Michael L J; Rutter, Guy A

    2002-01-01

    Glucose-responsive (GR) neurons from hypothalamic nuclei are implicated in the regulation of feeding and satiety. To determine the role of intracellular ATP in the closure of ATP-sensitive K+ (KATP) channels in these cells and associated glia, the cytosolic ATP concentration ([ATP]c) was monitored in vivo using adenoviral-driven expression of recombinant targeted luciferases and bioluminescence imaging. Arguing against a role for ATP in the closure of KATP channels in GR neurons, glucose (3 or 15 mm) caused no detectable increase in [ATP]c, monitored with cytosolic luciferase, and only a small decrease in the concentration of ATP immediately beneath the plasma membrane, monitored with a SNAP25–luciferase fusion protein. In contrast to hypothalamic neurons, hypothalamic glia responded to glucose (3 and 15 mm) with a significant increase in [ATP]c. Both neurons and glia from the cerebellum, a glucose-unresponsive region of the brain, responded robustly to 3 or 15 mm glucose with increases in [ATP]c. Further implicating an ATP-independent mechanism of KATP channel closure in hypothalamic neurons, removal of extracellular glucose (10 mm) suppressed the electrical activity of GR neurons in the presence of a fixed, high concentration (3 mm) of intracellular ATP. Neurons from both brain regions responded to 5 mm lactate (but not pyruvate) with an oligomycin-sensitive increase in [ATP]c. High levels of the plasma membrane lactate-monocarboxylate transporter, MCT1, were found in both cell types, and exogenous lactate efficiently closed KATP channels in GR neurons. These data suggest that (1) ATP-independent intracellular signalling mechanisms lead to the stimulation of hypothalamic neurons by glucose, and (2) these effects may be potentiated in vivo by the release of lactate from neighbouring glial cells. PMID:12381816

  14. A study on framework for assessing limiting condition for operation(LCOs) quantitatively in a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyung Min

    2003-02-15

    Limiting conditions for operations define the allowed outage times and the actions to be taken if the repair cannot be completed within the AOT. Typically the actions required are plant shutdown. In situations where the risk associated with the action, i.e., the risk of plant shutdown given a failure in the safety system, may be substantial, a strategy is needed to control the risk implications. In this study the strategy is evaluated quantitatively using a tool of system dynamics. The strategic actions associated with LCO needs to be assessed dynamically. System dynamics technique is easy enough to be applied for quantitative assessment of LCO, where the system dynamics is an important factor in evaluating operational modes of nuclear power plants. The review on the improvements of overly conservative technical specification has been performed in this study. The VENSIM tool has been applied to evaluate quantitatively the LCO of the auxiliary feed water systems in the Wol-Sung nuclear power plant. A value of core damage frequency in PSA is used as risk measure. The analysis of both full power operation and shutdown operation has been compared for the value of the CDF. This study shows a plot of LCO full power operating and shutdown risks in term of core damage probability for failure of AFWS (auxiliary feedwater system). Obtaining a lower risk level in a stable mode, compared to the LCO operation alternative, is the principal motivation of going to full power operational mode or shutdown operational mode. A time dependent framework developed in this study has been applied to assess the LCO of the example problem and it is shown that it is very flexible in that it can be applied to assess LCO quantitatively under any operational context of the technical specifications. This study contributes risk informed regulation to enhance and optimize the LCO of previous technical specifications for CANDU through developing framework in application proposed in this methodology.

  15. Reference Spent Fuel and Its Source Terms for a Design of Deep Geological Disposal System

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Dong Keun

    2005-12-15

    In this study, current status and future trend of domestic spent fuels were analyzed to propose reference spent nuclear fuel. And then, source terms needed for design of a deep geological disposal system were calculated using ORIGEN-ARP. The reference spent fuels selected based on assembly physical dimension, inventory projection, trend of initial enrichment of 235U, discharge burnup are as follows; The 17x17 Korean Optimized Fuel Assembly with initial enrichment of 4.0 wt.% 235U and discharge burnup of 45 GWD/MTU was adopted as a low-burnup representative fuel. For the high-burnup representative fuel, 16x16 Korean Standard Fuel Assembly with initial enrichment of 4.5 wt.% 235U and discharge burnup of 55 GWD/MTU was chosen. CANDU fuel with initial enrichment of 0.711 wt.% 235U and discharge burnup of 7.5 GWD/MTU was also considered. For these reference fuels, decay heat, radiation intensity and spectrum, nuclide concentration, and individual nuclide radioactivity were calculated using ORIGEN-ARP for a disposal system design. It is expected that the source terms estimated in this study will be applied to the disposal system development in the future.

  16. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy`s (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States [CIS]). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  17. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States (CIS)). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  19. Leeds-Lyon Symposium on Tribology ’The Tribology Design of Machine Elements’ (15th) 6-9 September 1988

    Science.gov (United States)

    1988-08-01

    In the Candu nuclear reactor, fuel bundles are moved into the channel by means of a ram connected to a ball screw. The screw is driven by a...lubricating film. Predominant mode of lubrication was probably mixed lubrication leading to high temperatures at the interface. Visual inspection of the...Velocity m/s Fig.4 ? -- SESSION XII (Paper (iv) - THURSDAY 8th SEPTEMBER WEAR PERFORMANCE OF MATERIALS FOR BALL SCREW AND SPLINE APPLICATIONS IN CANDU

  20. An experimental and simulation study of novel channel designs for open-cathode high-temperature polymer electrolyte membrane fuel cells

    DEFF Research Database (Denmark)

    Thomas, Sobi; Bates, Alex; Park, Sam

    2016-01-01

    A minimum balance of plant (BOP) is desired for an open-cathode high temperature polymer electrolyte membrane (HTPEM) fuel cell to ensure low parasitic losses and a compact design. The advantage of an open-cathode system is the elimination of the coolant plate and incorporation of a blower...... distribution through each cell. Design studies were carried out to increase power density. An experimental and simulation approach was carried out to design the novel open-cathode system. Two unique parallel serpentine flow designs were developed to yield a low pressure drop and uniform flow distribution, one...... without pins and another with pins. A five-cell stack was fabricated in the lab based on the new design. Performance and flow distribution studies revealed better performance, uniform flow distribution, and a reduced temperature gradient across the stack; improving overall system efficiency....

  1. Simulating thermal behavior of AECL's spent fuel dry storage system with CATHENA

    Energy Technology Data Exchange (ETDEWEB)

    Sabourin, G. [Atomic Energy of Canada Limited, Montreal, PQ (Canada)

    1998-07-01

    This paper documents the comparisons between CATHENA predictions and temperature measurements taken at the Gentilly-2 NPP spent fuel dry storage facility and in a mock--up of a storage basket placed inside a storage cylinder. It also presents CATHENA temperature predictions related to the storage of spent fuel in MACSTOR modules as planned for Ignalina NPP, Lithuania. CATHENA has been chosen because it can simulate many noncondensable gases including air and helium, and because of its great flexibility in the representation of the MACSTOR module geometry. The results of the simulations show good agreement with the experimental measurements. The two comparisons indicate that CATHENA can be used to simulate heat transfer from the fuel to the external air circuit of the spent fuel dry storage system. For the Ignalina MACSTOR module, containing RBMK fuel having higher heat release than typical CANDU fuel, CATHENA predicts that the maximum fuel temperature is expected to be around 240 deg C, giving an acceptable margin below the maximum allowed temperature of 300 deg C. In conclusion, this paper shows that the thermalhydraulic code CATHENA can accurately predict the thermal behavior AECL's air cooled spent fuel dry storage system. (author)

  2. Analysis of preliminary design concept of stainless steel container for disposal of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chun, K.S.; Ku, J.H.; Park, J.H.; Choi, J.W. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-02-01

    This report represents the structural, thermal and radiation shielding analysis of the basic concepts of the disposal container, that could accommodate PWR and CANDU fuels of which physical dimensions and shapes are quite different each other, with respect to the emplacement modes. Basic concepts of the disposal containers for the vertical horehole and the drift emplacement modes are proposed with their maximum allowable thermal loading. Appropriate thickness of the container to withstand the expected external pressure in the underground repository system was delivered by the structural analyses. The thermal analysis of the container containing spent fuels showed that the internal maximum temperatures of all container concepts did not reach the constraint values. Radiation dose rate from the container with 10cm thickness wall were also less than the established constraint value. (author). 9 refs., 33 figs., 12 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-04-15

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

  4. Spent fuel workshop'2002

    Energy Technology Data Exchange (ETDEWEB)

    Poinssot, Ch

    2002-07-01

    This document gathers the transparencies of the presentations given at the 2002 spent fuel workshop: Session 1 - Research Projects: Overview on the IN CAN PROCESSES European project (M. Cowper), Overview on the SPENT FUEL STABILITY European project (C. Poinssot), Overview on the French R and D project on spent fuel long term evolution, PRECCI (C. Poinssot); Session 2 - Spent Fuel Oxidation: Oxidation of uranium dioxide single crystals (F. Garrido), Experimental results on SF oxidation and new modeling approach (L. Desgranges), LWR spent fuel oxidation - effects of burn-up and humidity (B. Hanson), An approach to modeling CANDU fuel oxidation under dry storage conditions (P. Taylor); Session 3 - Spent Fuel Dissolution Experiments: Overview on high burnup spent fuel dissolution studies at FZK/INE (A. Loida), Results on the influence of hydrogen on spent fuel leaching (K. Spahiu), Leaching of spent UO{sub 2} fuel under inert and reducing conditions (Y. Albinsson), Fuel corrosion investigation by electrochemical techniques (D. Wegen), A reanalysis of LWR spent fuel flow through dissolution tests (B. Hanson), U-bearing secondary phases formed during fuel corrosion (R. Finch), The near-field chemical conditions and spent fuel leaching (D. Cui), The release of radionuclides from spent fuel in bentonite block (S.S. Kim), Trace actinide behavior in altered spent fuel (E. Buck, B. Hanson); Session 4 - Radiolysis Issues: The effect of radiolysis on UO{sub 2} dissolution determined from electrochemical experiments with {sup 238}Pu doped UO{sub 2} M. Stroess-Gascoyne (F. King, J.S. Betteridge, F. Garisto), doped UO{sub 2} studies (V. Rondinella), Preliminary results of static and dynamic dissolution tests with {alpha} doped UO{sub 2} in Boom clay conditions (K. Lemmens), Studies of the behavior of UO{sub 2} / water interfaces under He{sup 2+} beam (C. Corbel), Alpha and gamma radiolysis effects on UO{sub 2} alteration in water (C. Jegou), Behavior of Pu-doped pellets in brines

  5. Specification of ROP flux shape

    Energy Technology Data Exchange (ETDEWEB)

    Min, Byung Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Gray, A. [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1997-06-01

    The CANDU 9 480/SEU core uses 0.9% SEU (Slightly Enriched Uranium) fuel. The use f SEU fuel enables the reactor to increase the radial power form factor from 0.865, which is typical in current natural uranium CANDU reactors, to 0.97 in the nominal CANDU 9 480/SEU core. The difference is a 12% increase in reactor power. An additional 5% increase can be achieved due to a reduced refuelling ripple. The channel power limits were also increased by 3% for a total reactor power increase of 20%. This report describes the calculation of neutron flux distributions in the CANDU 9 480/SEU core under conditions specified by the C and I engineers. The RFSP code was used to calculate of neutron flux shapes for ROP analysis. Detailed flux values at numerous potential detector sites were calculated for each flux shape. (author). 6 tabs., 70 figs., 4 refs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-01-15

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

  7. Light Obscuration Particle Counter Fuel Contamination Limits

    Science.gov (United States)

    2015-10-08

    4) (5). The Army utilizes ASTM D4176 – Standard Test Method for Free Water and Particulate Contamination in Distillate Fuels (Visual Inspection ...where high in the 4µm (c) channel only. 78 samples where aviation fuels (JP-8 and F-24) that where high in the 6µm (c),14µm (c), and/or 30µm (c...AND USE OF LIQUID FUELS Charleston, South Carolina USA 4-8 October 2015 LIGHT OBSCURATION PARTICLE COUNTER FUEL CONTAMINATION LIMITS Joel

  8. MACSTOR{trademark}: Dry spent fuel storage for the nuclear power industry

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F.E.; Pattantyus, P. [AECL Candu, Montreal, Quebec (Canada); Hanson, A.S. [Transnuclear, Inc., Hawthorne, NY (United States)

    1993-12-31

    Safe storage of spent fuel has long been an area of critical concern for the nuclear power industry. As fuel pools fill up and re-racking possibilities become exhausted, power plant operators will find that they must ship spent fuel assemblies off-site or develop new on-site storage options. Many utility companies are turning to dry storage for their spent fuel assemblies. The MACSTOR (Modular Air-cooled Canister STORage) concept was developed with this in mind. Derived from AECL`s successful vertical loading, concrete silo program for storing CANDU nuclear spent fuel, MACSTOR was developed for light water reactor spent fuel and was subjected to full scale thermal testing. The MACSTOR Module is a monolithic, shielded concrete vault structure than can accommodate up to 24 spent fuel canisters. Each canister holds 12 PWR or 32 PWR previously cooled spent fuel assemblies with burn-up rates as high as 45,000 MWD/MTU. The structure is passively cooled by natural convection through an array of inlet and outlet gratings and galleries serving a central plenum where the (vertically) stored canisters are located. The canisters are continuously monitored by means of a pressure monitoring system developed by TNI. The MACSTOR system includes the storage module(s), an overhead gantry system for cask handling, a transfer cask for moving fuel from wet to dry storage and a cask transporter. The canister and transfer cask designs are based on Transnuclear transport cask designs and proven hot cell transfer cask technology, adapted to requirements for on-site spent fuel storage. This Modular Air Cooled System has a number of inherent advantages: efficient use of construction materials and site space; cooling is virtually impossible to impede; has the ability to monitor fuel confinement boundary integrity during storage; the fuel canisters may be used for both storage and transport and canisters utilize a flanged, ASME-III closure system that allows for easy inspection.

  9. In-membrane micro fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Omosebi, Ayokunle; Besser, Ronald

    2016-09-06

    An in-membrane micro fuel cell comprises an electrically-insulating membrane that is permissive to the flow of cations, such as protons, and a pair of electrodes deposited on channels formed in the membrane. The channels are arranged as conduits for fluids, and define a membrane ridge between the channels. The electrodes are porous and include catalysts for promoting the liberation of a proton and an electron from a chemical species and/or or the recombination of a proton and an electron with a chemical specie. The fuel cell may be provided a biosensor, an electrochemical sensor, a microfluidic device, or other microscale devices fabricated in the fuel cell membrane.

  10. FLASHPOINT - a tool to routinely calculate the heat load in the irradiated fuel bays

    Energy Technology Data Exchange (ETDEWEB)

    Vyskocil, E.; Morrison, C.; Gifford, E.; Inglot, A.; Kozlowski, K.; Gocmanac, M. [AMEC NSS, Reactor and Radiation Physics, Toronto, Ontario (Canada); Parlatan, Y. [Ontario Power Generation, Safety Analysis Improvement Project Dept., Pickering, Ontario (Canada); Alabasha, H. [Bruce Power, Nuclear Safety Analysis and Support, Toronto, Ontario (Canada)

    2013-07-01

    At the recommendation of the World Association of Nuclear Operators (WANO), a tool was developed as an enhancement of NuFLASH (Nuclear Fuel Location and Storage History) in order to routinely calculate the Irradiated Fuel Bay (IFB) heat load. It uses information stored in NuFLASH regarding the location and details of spent fuel bundle properties to calculate the decay power on a bundle by bundle basis and then sum the decay powers of all bundles in a particular IFB. FLASHPOINT employs a two-step approximation of the bundle irradiation history based on the record of the life cycle for each individual fuel bundle. The primary parameter affecting the decay power of any individual irradiated CANDU fuel bundle following its discharge from core is the period of time elapsed since the bundle last operated at power within the reactor. The remaining factors influencing the decay power of an individual fuel bundle concern the irradiation history of that bundle while in core. The accuracy of the FLASHPOINT methodology has been assessed primarily through comparison of results obtained using the two step history representation implemented in FLASHPOINT against results from a more detailed ORIGEN-S calculation of the decay heat based on the SORO power history for a randomly selected sample of bundles. The results for individual bundles and the aggregate group are presented and the accuracy of the two-step approximation is demonstrated to be acceptable. (author)

  11. A solid fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sonetaka, K.; Iketani, Y.; Nisino, A.; Takeuti, Y.

    1983-04-15

    A solid fuel is proposed which consists of cylindrical or prismatic pieces, whose surfaces (Pv) are equipped with greater than or equal to one channel, while the pieces themselves have greater than or equal to 1 through opening; the ratio of the total surface to the surface of the channels is within 95 to 5 to 60 to 40. The presence of the channels and the through openings facilitates the feeding of air to the surface of the fuel, providing in such a way, better ignition, the propagation of the flames and the completeness of combustion of the solid fuels based on carbonaceous materials. The optimal composition of the proposed fuel is: 70 to 95 percent carbonaceous base (coal, coke, charcoal, graphite, petroleum coke and so on), 1 to 25 percent desulfurizing additive (carbonate, hydroxide or nitrate of sodium, potassium, calcium or magnesium, 0 to 15 percent combustion accelerator (oxidizers: nitrates of potassium and barium, potassium perchlorate, oxides of magnesium, iron or manganese, aluminum powder and so on), 0 to 10 percent forming additive (bentonite, talc, kaolin and so on) and 0.5 to 5 percent binder (pitch, resin, pulp, cement and so on). The optimal characteristics of the combustion process are reached at a weight of a single piece of fuel of approximately 20 grams, a seeming density of less than or equal to 1.3 grams per cubic centimeter, a cross section surface area of 25 to 1,600 square millimeters and a filling density with combustion of less than or equal to 0.8 grams per cubic centimeter.

  12. Corrosion of used nuclear fuel in aqueous perchlorate and carbonate solutions

    Science.gov (United States)

    Shoesmith, D. W.; Sunder, S.; Bailey, M. G.; Miller, N. H.

    1996-01-01

    The corrosion of used fuel was investigated using electrodes constructed from fuel pins discharged from the Pickering, Bruce and Darlington CANDU reactors, and compared to the corrosion behaviour observed on unirradiated UO 2 and SIMFUEL. Experiments were carried out in solutions of NaClO 4 (pH˜ 9.5) in the presence and absence of (a) substantial concentrations of sodium carbonate, and (b) additional external gamma fields. Used fuel electrodes reached oxidizing corrosion potentials ( ECORR) rapidly compared with unirradiated UO 2 electrodes. However, optical and SEM examinations showed no evidence for rapid oxidative dissolution. This reaction, expected to be fast since high values of ECORR are observed, appears to be blocked by the accumulation of secondary phases in grain boundaries. The oxidation and dissolution behaviour of used fuel is determined predominantly by (i) the dose rate in solution near the fuel surface, (ii) the extent of burnup (which determines the degree of fission product doping), and (iii) the degree of non-stoichiometry.

  13. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    Energy Technology Data Exchange (ETDEWEB)

    Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

    1997-01-01

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs.

  14. Fossil Fuels.

    Science.gov (United States)

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  15. Fuel distribution

    Energy Technology Data Exchange (ETDEWEB)

    Tison, R.R.; Baker, N.R.; Blazek, C.F.

    1979-07-01

    Distribution of fuel is considered from a supply point to the secondary conversion sites and ultimate end users. All distribution is intracity with the maximum distance between the supply point and end-use site generally considered to be 15 mi. The fuels discussed are: coal or coal-like solids, methanol, No. 2 fuel oil, No. 6 fuel oil, high-Btu gas, medium-Btu gas, and low-Btu gas. Although the fuel state, i.e., gas, liquid, etc., can have a major impact on the distribution system, the source of these fuels (e.g., naturally-occurring or coal-derived) does not. Single-source, single-termination point and single-source, multi-termination point systems for liquid, gaseous, and solid fuel distribution are considered. Transport modes and the fuels associated with each mode are: by truck - coal, methanol, No. 2 fuel oil, and No. 6 fuel oil; and by pipeline - coal, methane, No. 2 fuel oil, No. 6 oil, high-Btu gas, medium-Btu gas, and low-Btu gas. Data provided for each distribution system include component makeup and initial costs.

  16. Analysis of the Reuse of Uranium Recovered from the Reprocessing of Commercial LWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    DelCul, Guillermo Daniel [ORNL; Trowbridge, Lee D [ORNL; Renier, John-Paul [ORNL; Ellis, Ronald James [ORNL; Williams, Kent Alan [ORNL; Spencer, Barry B [ORNL; Collins, Emory D [ORNL

    2009-02-01

    This report provides an analysis of the factors involved in the reuse of uranium recovered from commercial light-water-reactor (LWR) spent fuels (1) by reenrichment and recycling as fuel to LWRs and/or (2) by recycling directly as fuel to heavy-water-reactors (HWRs), such as the CANDU (registered trade name for the Canadian Deuterium Uranium Reactor). Reuse is an attractive alternative to the current Advanced Fuel Cycle Initiative (AFCI) Global Nuclear Energy Partnership (GNEP) baseline plan, which stores the reprocessed uranium (RU) for an uncertain future or attempts to dispose of it as 'greater-than-Class C' waste. Considering that the open fuel cycle currently deployed in the United States already creates a huge excess quantity of depleted uranium, the closed fuel cycle should enable the recycle of the major components of spent fuel, such as the uranium and the hazardous, long-lived transuranic (TRU) actinides, as well as the managed disposal of fission product wastes. Compared with the GNEP baseline scenario, the reuse of RU in the uranium fuel cycle has a number of potential advantages: (1) avoidance of purchase costs of 11-20% of the natural uranium feed; (2) avoidance of disposal costs for a large majority of the volume of spent fuel that is reprocessed; (3) avoidance of disposal costs for a portion of the depleted uranium from the enrichment step; (4) depending on the {sup 235}U assay of the RU, possible avoidance of separative work costs; and (5) a significant increase in the production of {sup 238}Pu due to the presence of {sup 236}U, which benefits somewhat the transmutation value of the plutonium and also provides some proliferation resistance.

  17. Mechanosensitive Channels

    Science.gov (United States)

    Martinac, Boris

    Living cells are exposed to a variety of mechanical stimuli acting throughout the biosphere. The range of the stimuli extends from thermal molecular agitation to potentially destructive cell swelling caused by osmotic pressure gradients. Cellular membranes present a major target for these stimuli. To detect mechanical forces acting upon them cell membranes are equipped with mechanosensitive (MS) ion channels. Functioning as molecular mechanoelectrical transducers of mechanical forces into electrical and/or chemical intracellular signals these channels play a critical role in the physiology of mechanotransduction. Studies of prokaryotic MS channels and recent work on MS channels of eukaryotes have significantly increased our understanding of their gating mechanism, physiological functions, and evolutionary origins as well as their role in the pathology of disease.

  18. Phased array ultrasonic technology (PAUT) contribution to detection and sizing of microbially influenced corrosion (MIC) of service water systems and shutdown coolers heat exchangers in OPG CANDU stations

    Energy Technology Data Exchange (ETDEWEB)

    Ciorau, P.; Pullia, L.; Hazelton, T., E-mail: peter.ciorau@opg.com, E-mail: lou.pullia@opg.com, E-mail: trek.hazelton@opg.com [Ontario Power Generation, Toronto, Ontario (Canada); Daks, W. [CAD WIRE, Markham, Ontario (Canada)

    2008-07-01

    Three PAUT techniques [linear scan - longitudinal waves, sector scan -longitudinal waves and sector scan-transverse waves] were developed and validated to assess the MIC attack in service water systems (SWS) and shutdown coolers heat-exchangers (SDC-HX) of Darlington and Pickering CANDU stations. PAUT employs linear array probes with a frequency between 4-12 MHz, depending on surface conditions, component geometry and MIC size/category to be detected. Examples from lab validation and field trials are presented. Based on field trials results, the techniques were optimized and new cal blocks were manufactured. It was demonstrated for mid-length pipes and for SDC-HX, the PAUT is the best technique compared with D-meter conventional UT and with guided waves. The expected field accuracy is about 0.5 mm (0.020{sup )} for large MIC attack. The ligament evaluation is technically achievable for colonies / pin holes located 2 mm under the outer surface. Improvements were identified and implemented for the next outages. (author)

  19. Fuel Cells

    DEFF Research Database (Denmark)

    Smith, Anders; Pedersen, Allan Schrøder

    2014-01-01

    Fuel cells have been the subject of intense research and development efforts for the past decades. Even so, the technology has not had its commercial breakthrough yet. This entry gives an overview of the technological challenges and status of fuel cells and discusses the most promising applications...... of the different types of fuel cells. Finally, their role in a future energy supply with a large share of fluctuating sustainable power sources, e.g., solar or wind, is surveyed....

  20. High specific power, direct methanol fuel cell stack

    Science.gov (United States)

    Ramsey, John C.; Wilson, Mahlon S.

    2007-05-08

    The present invention is a fuel cell stack including at least one direct methanol fuel cell. A cathode manifold is used to convey ambient air to each fuel cell, and an anode manifold is used to convey liquid methanol fuel to each fuel cell. Tie-bolt penetrations and tie-bolts are spaced evenly around the perimeter to hold the fuel cell stack together. Each fuel cell uses two graphite-based plates. One plate includes a cathode active area that is defined by serpentine channels connecting the inlet manifold with an integral flow restrictor to the outlet manifold. The other plate includes an anode active area defined by serpentine channels connecting the inlet and outlet of the anode manifold. Located between the two plates is the fuel cell active region.

  1. A transient fuel cell model to simulate HTPEM fuel cell impedance spectra

    DEFF Research Database (Denmark)

    Vang, Jakob Rabjerg; Andreasen, Søren Juhl; Kær, Søren Knudsen

    2011-01-01

    This paper presents a spatially resolved transient fuel cell model applied to the simulation of high temperature PEM fuel cell impedance spectra. The model is developed using a 2D finite volume method approach. The model is resolved along the channel and across the membrane. The model considers...

  2. Department of the Navy Supporting Data for Fiscal Year 1983 Budget Estimates Descriptive Summaries Submitted to Congress February 1982. Research, Development, Test and Evaluation, Navy. Book 3 of 3. Tactical Programs, Intelligence & Communications, Management & Support

    Science.gov (United States)

    1982-02-01

    34 ’,. scbheduled for fall 1982. Complete full-scale engineering development with primary emphiasis on actack capabilitied. Candu .- Bare .f Inspection and Survey...1277 S0829 Energy Conservation/Advanced- - ----- 1284 ,;038 Mobility. Fuels /Advanced...data ovet more channels at-greater range -than the current systems. -The Advanced Sonobuoy Communications Link system consists of an aircraft

  3. Metrology for Fuel Cell Manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Stocker, Michael [National Inst. of Standards and Technology, Gaithersburg, MD (United States); Stanfield, Eric [National Inst. of Standards and Technology, Gaithersburg, MD (United States)

    2015-02-04

    The project was divided into three subprojects. The first subproject is Fuel Cell Manufacturing Variability and Its Impact on Performance. The objective was to determine if flow field channel dimensional variability has an impact on fuel cell performance. The second subproject is Non-contact Sensor Evaluation for Bipolar Plate Manufacturing Process Control and Smart Assembly of Fuel Cell Stacks. The objective was to enable cost reduction in the manufacture of fuel cell plates by providing a rapid non-contact measurement system for in-line process control. The third subproject is Optical Scatterfield Metrology for Online Catalyst Coating Inspection of PEM Soft Goods. The objective was to evaluate the suitability of Optical Scatterfield Microscopy as a viable measurement tool for in situ process control of catalyst coatings.

  4. Brands & Channels

    Institute of Scientific and Technical Information of China (English)

    Alice Yang

    2009-01-01

    @@ "Brands" and "Channels" are the two most important things in Ku-Hai Chen's eyes when doing business with Main-land China. Ku-Hai Chen, Executive Director of the International Trade Institute of Taiwan External Trade Development Council (TAITRA), flies frequently between Chinese Taipei and Mainland China, and was in Beijing earlier this month for his seminar.

  5. Positron Channeling

    CERN Document Server

    Badikyan, Karen

    2016-01-01

    The possibility of channeling the low-energy relativistic positrons around separate crystallographic axes with coaxial symmetry of negative ions in some types of crystals is shown. The process of annihilation of positrons with electrons of medium was studied in detail.

  6. Gas Test Loop Booster Fuel Hydraulic Testing

    Energy Technology Data Exchange (ETDEWEB)

    Gas Test Loop Hydraulic Testing Staff

    2006-09-01

    The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3.

  7. 重水堆蒸汽旁路阀控制改进及优化%The Betterment of Design and Control in Candu CSDV’S

    Institute of Scientific and Technical Information of China (English)

    焦允站; 罗良盖; 高江波

    2016-01-01

    On the basis of introducing the Candu CSDVS function, The different problems are analyzed during design, commissioning, operation and Methods to solve these problems are analyzed in detail, And in view of the solution and results are introduced, Focus on logic control and the problems existing in the actuator control component were deeply analyzed, Put forward the feasible solution, Ensures that is properly during adjusting the valve in the control range in various working conditions, and still make the CSDVS control to the normal range, to avoid level fluctuations and unexpected shutdown.%在介绍重水堆蒸汽旁路系统功能的基础上,针对在设计、调试以及运行期间存在的不同问题进行了剖析,对这些问题的解决方法进行了详细的分析,并针对解决方案和成效进行了介绍。重点地对逻辑控制和执行机构控制组件存在的问题进行了较深入分析,提出切实可行的解决办法。确保了在机组各个工况下时能正常调节阀门在控制范围内,避免出现阀门控制响应异常,进而造成机组意外甩负荷和功率损失。

  8. Cofrentes EOC16B poolside measurements of channels from the three BWR vendors

    Energy Technology Data Exchange (ETDEWEB)

    Sedano, Pablo J. Garcia; Ayuela, Javier Iglesias [Iberdrola Ingenieria Construccion SAU, Veronica Anaures (Mexico); Albendea, Manuel [Iberdrola Generation S. A., Plaza Euskadi, 5 48009 BILBAO (Spain)

    2008-10-15

    As part of the EPRI Fuel Reliability Program, a fuel channel focus group was formed in 2002 to initiate measurements on irradiated BWR fuel channels. Fuel channels from GNF and AREVA have been measured in campaigns performed during 2004{approx}2007. Fuel channels designed and supplied by Westinghouse were of particular interest since no measurement information had been previously taken on modern Westinghouse channels operating on conventional loading pattern cycles, either in European or U.S. plants. Conventional loading pattern cycles are more susceptible to experience shadow corrosion induced bow since the fresh bundles are exposed to control blade influence early in life. During summer of 2007 extensive poolside measurements of a total of 180 fuel channels (24 SVEA-96 +/L, 68 SVEA-96 Optima-2, 36 GE-12, 42 GE-14 and 10 ATRIUM-10XP) have been performed by Westinghouse at Cofrentes NPP (Spanish BWR-6 operating on 24 month cycle strategy). This campaign has been co-sponsored by EPRI, Iberdrola and Westinghouse Sweden. Channels covering a range of exposure and control blade history were selected in order to determine the dependency of the channel deformation with those parameters. Channels with the most limiting conditions of exposure and control blade history were included. Channel bow, bulge and twist have been measured and fast neutron fluence calculations have been performed in order to determine the effects of neutron fluence gradient and shadow corrosion on the total channel deformation. Additionally channel oxide measurements have been performed on 20 channels from the three fuel vendors.The results indicate that channel bow and bulge remained at anticipated levels with no indication of significant channel bow due to shadow corrosion phenomenon. Destructive metallographic evaluations of samples taken from one cycle Westinghouse channels with high control blade exposure are underway at Studsvik hot cell facilities. These examinations will provide additional

  9. Reactor inspection and maintenance machine senses and homes in on reactor end fittings

    Energy Technology Data Exchange (ETDEWEB)

    Fell, R.G. [GE Canada Nuclear Products, Instrumentation and Control Engineering, Peterborough, Ontario (Canada)]. E-mail: bob.fell@cdnnuc.ge.com; Brown, R. [Ontario Power Generation, Inspection Services Div., Bowmanville, Ontario (Canada)

    2003-07-01

    The Universal Delivery Machine (UDM) is a new CANDU reactor maintenance tool that allows safe, timely, and cost-effective inspection and maintenance of fuel channels. The UDM must align precisely with reactor end-fittings in order to clamp onto fuel channels without applying excessive force. This alignment process is called fine homing. This paper describes the UDM instrumentation and control design features used in the fine homing process. (author)

  10. Channel Power in Multi-Channel Environments

    NARCIS (Netherlands)

    M.G. Dekimpe (Marnik); B. Skiera (Bernd)

    2004-01-01

    textabstractIn the literature, little attention has been paid to instances where companies add an Internet channel to their direct channel portfolio. However, actively managing multiple sales channels requires knowing the customers’ channel preferences and the resulting channel power. Two key compon

  11. Characterization of Hydrogen Content in ZIRCALOY-4 Nuclear Fuel Cladding

    Science.gov (United States)

    Pfeif, E. A.; Lasseigne, A. N.; Krzywosz, K.; Mader, E. V.; Mishra, B.; Olson, D. L.

    2010-02-01

    Assessment of hydrogen uptake of underwater nuclear fuel clad and component materials will enable improved monitoring of fuel health. Zirconium alloys are used in nuclear reactors as fuel cladding, fuel channels, guide tubes and spacer grids, and are available for inspection in spent fuel pools. With increasing reactor exposure zirconium alloys experience hydrogen ingress due to neutron interactions and water-side corrosion that is not easily quantified without destructive hot cell examination. Contact and non-contact nondestructive techniques, using Seebeck coefficient measurements and low frequency impedance spectroscopy, to assess the hydrogen content and hydride formation within zircaloy 4 material that are submerged to simulate spent fuel pools are presented.

  12. Nonlinear channelizer

    Science.gov (United States)

    In, Visarath; Longhini, Patrick; Kho, Andy; Neff, Joseph D.; Leung, Daniel; Liu, Norman; Meadows, Brian K.; Gordon, Frank; Bulsara, Adi R.; Palacios, Antonio

    2012-12-01

    The nonlinear channelizer is an integrated circuit made up of large parallel arrays of analog nonlinear oscillators, which, collectively, serve as a broad-spectrum analyzer with the ability to receive complex signals containing multiple frequencies and instantaneously lock-on or respond to a received signal in a few oscillation cycles. The concept is based on the generation of internal oscillations in coupled nonlinear systems that do not normally oscillate in the absence of coupling. In particular, the system consists of unidirectionally coupled bistable nonlinear elements, where the frequency and other dynamical characteristics of the emergent oscillations depend on the system's internal parameters and the received signal. These properties and characteristics are being employed to develop a system capable of locking onto any arbitrary input radio frequency signal. The system is efficient by eliminating the need for high-speed, high-accuracy analog-to-digital converters, and compact by making use of nonlinear coupled systems to act as a channelizer (frequency binning and channeling), a low noise amplifier, and a frequency down-converter in a single step which, in turn, will reduce the size, weight, power, and cost of the entire communication system. This paper covers the theory, numerical simulations, and some engineering details that validate the concept at the frequency band of 1-4 GHz.

  13. Computer-Based Procedure Systems: Technical Basis and Human Factors Review Guidance

    Science.gov (United States)

    2000-03-01

    computer-assisted system for fuel reloading while at power was designed for CANDU NPPs (Gertman et al., 1994). AECL has several aids under...for visualizing the interior of a reactor fuel channel to support the removal of stuck fuel bundles. This system is envisioned as a training aid...and vendor documents and correspondence; NRC correspondence and internal memoranda; bulletins and information no- tices; inspection and

  14. Recent findings on the oxidation of UO{sub 2} fuel under nominally dry storage conditions

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, P.; McEachern, R.J.; Sunder, S.; Wasywich, K.M.; Miller, N.H.; Wood, D.D

    1995-07-01

    This paper is an overview of fuel-storage demonstration experiments, supporting research on UO{sub 2} oxidation, and associated model development, in progress at AECL's Whiteshell Laboratories. The work is being performed to determine the time/temperature limits for safe storage of irradiated CANDU fuel in dry air. The most significant recent experimental finding has been the detection of small quantities of U{sub 3}O{sub 8}, formed over periods of one to several years in a variety of experiments at 150-170 deg C. Another important trading is the slight suppression of U{sub 3}O{sub 8} formation in SIMFUEL and other doped U0{sub 2} formulations. The development of a nucleation-and-growth model for U{sub 3}O{sub 8} formation is discussed, along with available activation energy data. These provide a basis for predicting U{sub 3}O{sub 8} formation rates under dry-storage conditions, and hence optimizing fuel storage strategies. (author)

  15. Fuel cells

    Directory of Open Access Journals (Sweden)

    D. N. Srivastava

    1962-05-01

    Full Text Available The current state of development of fuel cells as potential power sources is reviewed. Applications in special fields with particular reference to military requirements are pointed out.

  16. Future Fuels

    Science.gov (United States)

    2006-04-01

    Storage Devices, Fuel Management, Gasification, Fischer-Tropsch, Syngas , Hubberts’s Peak UNCLAS UNCLAS UNCLAS UU 80 Dr. Sujata Millick (703) 696...prices ever higher, and perhaps lead to intermittent fuel shortages as production fluctuates. Clearly, this competition for resources also provides oil...producers multiple options for selling their products, and raises the possibility that the US could face shortages resulting from shifts in

  17. Deep geological disposal system development; mechanical structural stability analysis of spent nuclear fuel disposal canister under the internal/external pressure variation

    Energy Technology Data Exchange (ETDEWEB)

    Kwen, Y. J.; Kang, S. W.; Ha, Z. Y. [Hongik University, Seoul (Korea)

    2001-04-01

    This work constitutes a summary of the research and development work made for the design and dimensioning of the canister for nuclear fuel disposal. Since the spent nuclear fuel disposal emits high temperature heats and much radiation, its careful treatment is required. For that, a long term(usually 10,000 years) safe repository for spent fuel disposal should be securred. Usually this repository is expected to locate at a depth of 500m underground. The canister construction type introduced here is a solid structure with a cast iron insert and a corrosion resistant overpack, which is designed for spent nuclear fuel disposal in a deep repository in the crystalline bedrock, which entails an evenly distributed load of hydrostatic pressure from undergroundwater and high pressure from swelling of bentonite buffer. Hence, the canister must be designed to withstand these high pressure loads. Many design variables may affect the structural strength of the canister. In this study, among those variables array type of inner baskets and thicknesses of outer shell and lid and bottom are tried to be determined through the mechanical linear structural analysis, thicknesses of outer shell is determined through the nonlinear structural analysis, and the bentonite buffer analysis for the rock movement is conducted through the of nonlinear structural analysis Also the thermal stress effect is computed for the cast iron insert. The canister types studied here are one for PWR fuel and another for CANDU fuel. 23 refs., 60 figs., 23 tabs. (Author)

  18. The PACo channel strategy; La strategie du reseau PACo

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    The PACo channel was created in june 1999 to answer government demands in the domain of new energies, to stimulate the new technologies development and to participate to enterprises growth. The first year was devoted in particular to the fuel cell and the hydrogen fuel. This document presents an analysis of the first year of program management. (A.L.B.)

  19. MACSTOR, an on-site, dry, spent-fuel storage system developed by AECL for use by U. S. utilities

    Energy Technology Data Exchange (ETDEWEB)

    Durante, R.; Feinroth, H.; Pattanyus, P. (AECL Technologies, Bethesda, MD (United States))

    1992-01-01

    The continuing delay in the U.S. Department of Energy's Yucca Mountain and monitored retrievable storage spent-fuel disposal and storage programs has prompted U.S. utilities to consider expanding on-site storage of spent reactor fuel. Long-term, on-site storage has certain advantages to U.S. utilities since it eliminates the need for costly and difficult shipping and puts control of the spent fuel completely under the direction of the owner-utility. AECL Technologies (AECL), through its research company and Canada deuterium uranium (CANDU) engineering services division, has been developing on-site storage for Canadian heavy water nuclear plants for almost 20 yr. AECL has developed a design for a dry storage unit, designated MACSTOR (modular air-cooled storage), that can accommodate U.S. light water reactor (LWR) fuel elements and could become a candidate for the U.S. market. This paper describes MACSTOR and its evolution from the original silos and CANSTOR system that was developed and used in Canada. These systems are subject to regulatory controls by the Atomic Energy Control Board of Canada and have proven to be safe, convenient, and cost effective.

  20. Numerical simulation of liquid water and gas flow in a channel and a simplified gas diffusion layer model of polymer electrolyte membrane fuel cells using the lattice Boltzmann method

    OpenAIRE

    Tabe, Yutaka; Lee, Yongju; Chikahisa, Takemi; Kozakai, Masaya

    2009-01-01

    Numerical simulations using the lattice Boltzmann method (LBM) are developed to elucidate the dynamic behavior of condensed water and gas flow in a polymer electrolyte membrane (PEM) fuel cell. Here, the calculation process of the LBM simulation is improved to extend the simulation to a porous medium like a gas diffusion layer (GDL), and a stable and reliable simulation of two-phase flow with large density differences in the porous medium is established. It is shown that dynamic capillary fin...

  1. Micro PEM Fuel Cells and Stacks

    Institute of Scientific and Technical Information of China (English)

    Shou-shing; Hsieh

    2007-01-01

    1 Results The effects of different operating parameters on micro proton exchange membrane (PEM) fuel cell performance were experimentally studied for three different flow field configurations (interdigitated,mesh,and serpentine).Experiments with different cell operating temperatures and different backpressures on the H2 flow channels,as well as various combinations of these parameters,have been conducted for three different flow geometries.The micro PEM fuel cells were designed and fabricated in-house t...

  2. HANARO core channel flow-rate measurement

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Heon Il; Chae, Hee Tae; Im, Don Soon; Kim, Seon Duk [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    HANARO core consists of 23 hexagonal flow tubes and 16 cylindrical flow tubes. To get the core flow distribution, we used 6 flow-rate measuring dummy fuel assemblies (instrumented dummy fuel assemblies). The differential pressures were measured and converted to flow-rates using the predetermined relationship between AP and flow-rate for each instrumented dummy fuel assemblies. The flow-rate for the cylindrical flow channels shows +-7% relative errors and that for the hexagonal flow channels shows +-3.5% relative errors. Generally the flow-rates of outer core channels show smaller values compared to those of inner core. The channels near to the core inlet pipe and outlet pipes also show somewhat lower flow-rates. For the lower flow channels, the thermal margin was checked by considering complete linear power histories. From the experimental results, the gap flow-rate was estimated to be 49.4 kg/s (cf. design flow of 50 kg/s). 15 tabs., 9 figs., 10 refs. (Author) .new.

  3. Fuel cells:

    DEFF Research Database (Denmark)

    Sørensen, Bent

    2013-01-01

    A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil and nucl......A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil...

  4. Solar fuels

    Energy Technology Data Exchange (ETDEWEB)

    Bolton, J.R.

    1978-11-17

    The paper is concerned with (1) the thermodynamic and kinetic limits for the photochemical conversion and storage of solar energy as it is received on the earth's surface, and (2) the evaluation of a number of possible photochemical reactions with particular emphasis on the production of solar hydrogen from water. Procedures for generating hydrogen fuel are considered. Topics examined include the general requirements for a fuel-generation reaction, the photochemical reaction, limits on the conversion of light energy to chemical energy, an estimate of chemical storage efficiency, and the water decomposition reaction.

  5. A Comparison of Materials Issues for Cermet and Graphite-Based NTP Fuels

    Science.gov (United States)

    Stewart, Mark E.; Schnitzler, Bruce G.

    2013-01-01

    This paper compares material issues for cermet and graphite fuel elements. In particular, two issues in NTP fuel element performance are considered here: ductile to brittle transition in relation to crack propagation, and orificing individual coolant channels in fuel elements. Their relevance to fuel element performance is supported by considering material properties, experimental data, and results from multidisciplinary fluid/thermal/structural simulations. Ductile to brittle transition results in a fuel element region prone to brittle fracture under stress, while outside this region, stresses lead to deformation and resilience under stress. Poor coolant distribution between fuel element channels can increase stresses in certain channels. NERVA fuel element experimental results are consistent with this interpretation. An understanding of these mechanisms will help interpret fuel element testing results.

  6. Multiply manifolded molten carbonate fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Krumpelt, M.; Roche, M.F.; Geyer, H.K.; Johnson, S.A.

    1994-08-01

    This study consists of research and development activities related to the concept of a molten carbonate fuel cell (MCFC) with multiple manifolds. Objective is to develop an MCFC having a higher power density and a longer life than other MCFC designs. The higher power density will result from thinner gas flow channels; the extended life will result from reduced temperature gradients. Simplification of the gas flow channels and current collectors may also significantly reduce cost for the multiply manifolded MCFC.

  7. Composition heterogeneity analysis for DUPIC fuel(I) - Statistical analysis

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-08-01

    The fuel composition heterogeneity effect on reactor performance parameters was assessed by refueling simulations for three DUPIC fuel options of fuel composition heterogeneity control: the fissile content adjustment, the reactivity control by slightly enriched and depleted uranium, and the reactivity control by natural uranium. For each DUPIC fuel option, the simulations were performed using 30 heterogeneous fuel types which were determined by the agglomerative hierarchical clustering method. The heterogeneity effect was considered during the refueling simulation by randomly selecting fuel types for the refueling operation. The refueling simulations of the heterogeneous core have shown that the key performance parameters such as the maximum channel power (MCP), maximum bundle power (MBP), and channel power peaking factor (CPPF) are close to those of the core that has single fuel type. For the three DUPIC fuel options, the uncertainties of MCP, MBP, and CPPF due to the fuel composition heterogeneity are less than 0.6, 1.5 and 0.8%, respectively, including the uncertainty of the group-average fuel property. This study has shown that the three DUPIC fuel options reduces the composition heterogeneity effectively and the zone power control system has a sufficient margin to adjust the perturbations cased by the fuel composition heterogeneity. 15 refs., 28 figs.,10 tabs. (Author)

  8. Transport fuel

    DEFF Research Database (Denmark)

    Ronsse, Frederik; Jørgensen, Henning; Schüßler, Ingmar

    2014-01-01

    Worldwide, the use of transport fuel derived from biomass increased four-fold between 2003 and 2012. Mainly based on food resources, these conventional biofuels did not achieve the expected emission savings and contributed to higher prices for food commod - ities, especially maize and oilseeds...

  9. Fuel Cells

    Science.gov (United States)

    Hawkins, M. D.

    1973-01-01

    Discusses the theories, construction, operation, types, and advantages of fuel cells developed by the American space programs. Indicates that the cell is an ideal small-scale power source characterized by its compactness, high efficiency, reliability, and freedom from polluting fumes. (CC)

  10. Experimental study and comparison of various designs of gas flow fields to PEM fuel cells and cell stack performance

    Directory of Open Access Journals (Sweden)

    Hong eLiu

    2014-01-01

    Full Text Available In this study, a significant number of experimental tests to PEM fuel cells were conducted to investigate the effect of gas flow fields on fuel cell performance. Graphite plates with various flow field or flow channel designs, from literature survey and also novel designs by the authors, were used for the PEM fuel cell assembly. The fabricated fuel cells all have an effective membrane area of 23.5 cm2. The results showed that the serpentine flow channel design is still favorable, giving the best single fuel cell performance amongst all the studied flow channel designs. A novel symmetric serpentine flow field was proposed for relatively large size fuel cell application. Four fuel cell stacks each including four cells were assembled using different designs of serpentine flow channels. The output power performances of fuel cell stacks were compared and the novel symmetric serpentine flow field design is recommended for its very good performance.

  11. A Review of Thorium Utilization as an option for Advanced Fuel Cycle--Potential Option for Brazil in the Future

    Energy Technology Data Exchange (ETDEWEB)

    Maiorino, J.R.; Carluccio, T.

    2004-10-03

    Since the beginning of Nuclear Energy Development, Thorium was considered as a potential fuel, mainly due to the potential to produce fissile uranium 233. Several Th/U fuel cycles, using thermal and fast reactors were proposed, such as the Radkwoski once through fuel cycle for PWR and VVER, the thorium fuel cycles for CANDU Reactors, the utilization in Molten Salt Reactors, the utilization of thorium in thermal (AHWR), and fast reactors (FBTR) in India, and more recently in innovative reactors, mainly Accelerator Driven System, in a double strata fuel cycle. All these concepts besides the increase in natural nuclear resources are justified by non proliferation issues (plutonium constrain) and the waste radiological toxicity reduction. The paper intended to summarize these developments, with an emphasis in the Th/U double strata fuel cycle using ADS. Brazil has one of the biggest natural reserves of thorium, estimated in 1.2 millions of tons of ThO{sub 2}, as will be reviewed in this paper, and therefore R&D programs would be of strategically national interest. In fact, in the past there was some projects to utilize Thorium in Reactors, as the ''Instinto/Toruna'' Project, in cooperation with France, to utilize Thorium in Pressurized Heavy Water Reactor, in the mid of sixties to mid of seventies, and the thorium utilization in PWR, in cooperation with German, from 1979-1988. The paper will review these initiatives in Brazil, and will propose to continue in Brazil activities related with Th/U fuel cycle.

  12. A transient model to simulate HTPEM fuel cell impedance spectra

    DEFF Research Database (Denmark)

    Vang, Jakob Rabjerg; Andreasen, Søren Juhl; Kær, Søren Knudsen

    2012-01-01

    This paper presents a spatially resolved transient fuel cell model applied to the simulation of high temperature PEM fuel cell impedance spectra. The model is developed using a 2D finite volume method approach. The model is resolved along the channel and across the membrane. The model considers...

  13. Analog information and the Canadian concept for disposal of nuclear fuel waste

    Energy Technology Data Exchange (ETDEWEB)

    Cramer, J.J. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada)

    1996-07-01

    AECL, with support from Ontario Hydro under auspices of the Candu Owners Group, has assessed a concept for the safe disposal of nuclear fuel waste in Canada. The disposal concept is to place nuclear fuel waste in corrosion-resistant containers and emplace the containers with sealing materials in an engineered vault at depths of 500 to 1000m in plutonic rock of the Canadian Shield. Humans and the environment would be protected from contaminants in the waste by several barriers; the waste itself, the container, the sealing materials, and the rock. This disposal concept permits a great deal of flexibility in its implementation, which means that a wide range of circumstances could be accommodated. Studies of natural analogues provide important information for evaluating and improving our knowledge and understanding of the disposal concept. Analogue information is used to develop the scenarios and conceptual models, to provide input to databases, and to test models, thereby enhancing the level of confidence in the safety predictions from the assessment models. In addition, natural analogues are valuable illustrative tools when presenting information on the disposal concept to the non-expert and the public.

  14. Development of linear sensitivity matrix method for fast evaluation of CANDU refuelling schemes%用于CANDU堆换料方案快速评价的线性敏感矩阵方法

    Institute of Scientific and Technical Information of China (English)

    王文聪; 刘宇轩; 冯进军; 王军; 樊申; 张少泓

    2009-01-01

    为了实现CANDU堆换料方案的快速评价,本文提出了线性敏感矩阵方法(LSM).该方法基于各种扰动引起的堆芯状态变化互不干涉且堆芯对扰动的响应与扰动量成正比的假设,借助于事先针对参考堆芯形成的敏感矩阵,主要的堆芯参数经简单代数运算就可得到,无需耗时的三维扩散方程求解.秦山三期1号机组部分运行历史检验结果表明,该方法不但可大幅提高换料方案评价的速度,而且还具有较高的精度,能满足现场工程应用的要求.%In order to develop a numerical tool for the fast evaluation of CANDU refuelling schemes, a Linear Sensitivity Matrix method (LSM) is developed. It assumes that all the effects caused by various perturbations to the core state are independent to each other and the core response to a perturbation is proportional to its magnitude. In this way, the main core parameters of a refuelling scheme can be derived by simple algebraic operations with the use of pre-calculated sensitivity matrices, without resorting to the complicated and time-consuming 3D core calculation. Verification against the QinShan CANDU reactor operation history demonstrates that LSM is capable of generating accurate results and running very fast for evaluating a refuelling scheme.

  15. Local Fuel Rod Crud Prediction Tool Applications

    Energy Technology Data Exchange (ETDEWEB)

    Krammen, Michael A.; Karoutas, Zeses E.; Wang, Guoqiang; Young, Michael Y

    2009-06-15

    response to two CILC fuel failures observed in a reactor plant where fuel rod crudding was not initially of concern due to negligible predicted fuel rod steaming on a fuel assembly sub-channel scale. The fuel rod crud observed in the reactor cycle with the fuel rod CILC failures was very localized, but was heaviest on those fuel rods with relatively higher fuel rod duty. These indications led to development of the more locally detailed predictive capability. And, based on the observed behavior, guideline limits have been established by benchmarking the methodology to the fuel rod crud induced fuel failures. The guideline limits are used in designing fuel managements. Application of these tools in subsequent fuel management design for later reactor cycles in both the plant where the CILC fuel failures occurred and in its sister plant with similar operating characteristics have avoided a recurrence of the CILC fuel failure. These tools were also used when a new fuel design with mixing vane grids was introduced in two plants previously fueled with non-mixing vane grids. The predictive tools account for the thermal hydraulic transition core effects. Interestingly, the plant with the generally higher fuel duty, a plant that had experienced Crud Induced Power Shift (CIPS) in earlier cycles, is predicted to easily meet the CIPS and CILC guidelines for the transition and following cycle. While the other plant, which has not experienced CIPS in earlier cycles, is predicted to be operating close to the CILC guideline limits in the transition cycle. The higher duty plant is predicted to have appreciable fuel rod surface area that is steaming over the reactor cycle, while the lower duty plant is predicted to have relatively little fuel rod surface area in steaming. The interpretation is that with a relatively similar crud source in the coolant, a smaller steaming surface area may act as a stronger sink for the available crud, resulting in locally thicker crud. This is a similar

  16. Nuclear fuel cycle assessment of India: A technical study for U.S.-India cooperation

    Science.gov (United States)

    Krishna, Taraknath Woddi Venkat

    breeder core concept involving the CANDU core design. The end-of-life fuel characteristics evolved from the designed fuel composition is proliferation resistant and economical in integrating this technology into the Indian nuclear fuel cycle. Furthermore, it is shown that the separation of the military and civilian components of the Indian fuel cycle can be facilitated through the implementation of such a system.

  17. Air-cooled, hydrogen-air fuel cell

    Science.gov (United States)

    Shelekhin, Alexander B. (Inventor); Bushnell, Calvin L. (Inventor); Pien, Michael S. (Inventor)

    1999-01-01

    An air-cooled, hydrogen-air solid polymer electrolyte (SPE) fuel cell with a membrane electrode assembly operatively associated with a fluid flow plate having at least one plate cooling channel extending through the plate and at least one air distribution hole extending from a surface of the cathode flow field into the plate cooling channel.

  18. Thermal behavior in the cracking reaction zone of scramjet cooling channels at different channel aspect ratios

    Science.gov (United States)

    Zhang, Silong; Feng, Yu; Jiang, Yuguang; Qin, Jiang; Bao, Wen; Han, Jiecai; Haidn, Oskar J.

    2016-10-01

    To study the thermal behavior in the cracking reaction zone of regeneratively cooled scramjet cooling channels at different aspect ratios, 3-D model of fuel flow in terms of the fuel's real properties and cracking reaction is built and validated through experiments. The whole cooling channel is divided into non-cracking and cracking reaction zones. Only the cracking reaction zone is studied in this article. The simulation results indicate that the fuel conversion presents a similar distribution with temperature because the fuel conversion in scramjet cooling channels is co-decided by the temperature and velocity but the temperature plays the dominate role. For the cases given in this paper, increasing the channel aspect ratio will increase the pressure drop and it is not beneficial for reducing the wall temperature because of the much severer thermal stratification, larger conversion non-uniformity, the corresponding M-shape velocity profile which will cause local heat transfer deterioration and the decreased chemical heat absorption. And the decreased chemical heat absorption caused by stronger temperature and conversion non-uniformities is bad for the utilization of chemical heat sink, chemical recuperation process and the ignition performance.

  19. Fuel control system for dual fuel engines

    Energy Technology Data Exchange (ETDEWEB)

    Helmich, M.J.; Ryan, W.P.; Marvin, D.H.

    1987-11-24

    A fuel governing system for an engine adapted for operation on a first fuel and a second fuel is described comprising: a first fuel governing system including a spontaneous motion metering means; and a second fuel governing system, the second fuel governing system further comprising: means for providing a first signal indicative of position of the first fuel metering means, which signal approximates total load on the engine, means for providing a second signal of the selected percentage of first fuel relative to total load, means for controlling flow of the second fuel to the engine, which flow causes reflective displacement of the first fuel metering means, means for determining the difference between the first signal and the second signal, which difference is indicative of distance the first fuel metering means must be moved to attain the selected percentage of first fuel relative to total load, and means for causing operation of the means for controlling flow of the second fuel to the engine to cause displacement of the first fuel metering means equal to the distance the first fuel metering means must be moved to attain the selected percentage of first fuel relative to total load.

  20. Thermal Aspects of Using Alternative Nuclear Fuels in Supercritical Water-Cooled Reactors

    Science.gov (United States)

    Grande, Lisa Christine

    A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT)-type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties of reactor coolant, fuel sheath and fuel. Uniform and non-uniform AHFPs for average channel power were applied to a variety of alternative fuels (mixed oxide, thorium dioxide, uranium dicarbide, uranium nitride and uranium carbide) enclosed in an Inconel-600 43-element bundle. The results depict bulk-fluid, outer-sheath and fuel-centreline temperature profiles together with the Heat Transfer Coefficient (HTC) profiles along the heated length of fuel channel. The objective is to identify the best options in terms of fuel, sheath material and AHFPS in which the outer-sheath and fuel-centreline temperatures will be below the accepted temperature limits of 850°C and 1850°C respectively. The 43-element Inconel-600 fuel bundle is suitable for SCWR use as the sheath-temperature design limit of 850°C was maintained for all analyzed cases at average channel power. Thoria, UC2, UN and UC fuels for all AHFPs are acceptable since the maximum fuel-centreline temperature does not exceed the industry accepted limit of 1850°C. Conversely, the fuel-centreline temperature limit was exceeded for MOX at all AHFPs, and UO2 for both cosine and downstream-skewed cosine AHFPs. Therefore, fuel-bundle modifications are required for UO2 and MOX to be feasible nuclear fuels for SCWRs.

  1. Stationary Liquid Fuel Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Won Sik [Purdue Univ., West Lafayette, IN (United States); Grandy, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Boroski, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, Lubomir [Argonne National Lab. (ANL), Argonne, IL (United States); Johnson, Terry [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-09-30

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

  2. Digital motion control for Ontario Power Generation's universal delivery machine

    Energy Technology Data Exchange (ETDEWEB)

    Murray, D.W. [GE Canada Nuclear Products, Peterborough, Ontario (Canada)]. E-mail: David.Murray@cdnnuc.ge.com; Roberts, A. [Ontario Power Generation, Inc., Inspection Services Div., Tiverton, Ontario (Canada)]. E-mail: anthony.roberts@ontariopowergeneration.com

    2003-07-01

    The Universal Delivery Machine (UDM) has thirteen axes of motion with a variety of precision positioning requirements designed to deliver reactor channel maintenance tooling into the CANDU fuel channels. These axes of motion provide the UDM with the capability to home and lock to the fuel channels, remove and replace the closure and shield plugs, advance, retract and rotate rams for fuel removal and for the deployment of the inspection and maintenance tooling. This paper describes the digital motion control system used on UDM. It will review the benefits of a digital motion control system while looking at the drive components used on the UDM. (author)

  3. Characterizing high-temperature deformation of internally heated nuclear fuel element simulators

    Energy Technology Data Exchange (ETDEWEB)

    Belov, A.I.; Fong, R.W.L.; Leitch, B.W.; Nitheanandan, T.; Williams, A., E-mail: alexander.belov@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    The sag behaviour of a simulated nuclear fuel element during high-temperature transients has been investigated in an experiment utilizing an internal indirect heating method. The major motivation of the experiment was to improve understanding of the dominant mechanisms underlying the element thermo-mechanical response under loss-of-coolant accident conditions and to obtain accurate experimental data to support development of 3-D computational fuel element models. The experiment was conducted using an electrically heated CANDU fuel element simulator. Three consecutive thermal cycles with peak temperatures up to ≈1000 {sup o}C were applied to the element. The element sag deflections and sheath temperatures were measured. On heating up to 600 {sup o}C, only minor lateral deflections of the element were observed. Further heating to above 700 {sup o}C resulted in an element multi-rate creep and significant permanent bow. Post-test visual and X-ray examinations revealed a pronounced necking of the sheath at the pellet-to-pellet interface locations. A wall thickness reduction was detected in the necked region that is interpreted as a sheath longitudinal strain localization effect. The sheath cross-sectioning showed signs of a 'hard' pellet-cladding interaction due to the applied cycles. A 3-D model of the experiment was generated using the ANSYS finite element code. As a fully coupled thermal mechanical simulation is computationally expensive, it was deemed sufficient to use the measured sheath temperatures as a boundary condition, and thus an uncoupled mechanical simulation only was conducted. The ANSYS simulation results match the experiment sag observations well up to the point at which the fuel element started cooling down. (author)

  4. Aviation fuels outlook

    Science.gov (United States)

    Momenthy, A. M.

    1980-01-01

    Options for satisfying the future demand for commercial jet fuels are analyzed. It is concluded that the most effective means to this end are to attract more refiners to the jet fuel market and encourage development of processes to convert oil shale and coal to transportation fuels. Furthermore, changing the U.S. refineries fuel specification would not significantly alter jet fuel availability.

  5. Mobile radio channels

    CERN Document Server

    Pätzold, Matthias

    2011-01-01

    Providing a comprehensive overview of the modelling, analysis and simulation of mobile radio channels, this book gives a detailed understanding of fundamental issues and examines state-of-the-art techniques in mobile radio channel modelling. It analyses several mobile fading channels, including terrestrial and satellite flat-fading channels, various types of wideband channels and advanced MIMO channels, providing a fundamental understanding of the issues currently being investigated in the field. Important classes of narrowband, wideband, and space-time wireless channels are explored in deta

  6. Channel nut tool

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Marvin

    2016-01-12

    A method, system, and apparatus for installing channel nuts includes a shank, a handle formed on a first end of a shank, and an end piece with a threaded shaft configured to receive a channel nut formed on the second end of the shaft. The tool can be used to insert or remove a channel nut in a channel framing system and then removed from the channel nut.

  7. Analysis Of The Effect Of Flow Channel Width On The Performance Of PEMFC

    OpenAIRE

    Elif Eker; İmdat Taymaz

    2013-01-01

    In this work, it was analysed the effect of different channel width on performance of PEM fuel cell. Current density were measured on the single cells of parallel flow fields that has 25 cm² active layer, using three different kinds of channel width. The cell width and the channel height remain constant.The results show that increasing the channel width while the cell width remains constant decreases the current density.

  8. Analysis Of The Effect Of Flow Channel Width On The Performance Of PEMFC

    OpenAIRE

    Eker, Elif; Taymaz, İmdat

    2013-01-01

    In this work, it was analysed the effect of different channel width on performance of PEM fuel cell. Current density were measured on the single cells of parallel flow fields that has 25 cm² active layer, using three different kinds of channel width. The cell width and the channel height remain constant. The results show that increasing the channel width while the cell width remains constant decreases the current density.

  9. Fuel injection nozzle and method of manufacturing the same

    Energy Technology Data Exchange (ETDEWEB)

    Monaghan, James Christopher; Johnson, Thomas Edward; Ostebee, Heath Michael

    2017-02-21

    A fuel injection head for use in a fuel injection nozzle comprises a monolithic body portion comprising an upstream face, an opposite downstream face, and a peripheral wall extending therebetween. A plurality of pre-mix tubes are integrally formed with and extend axially through the body portion. Each of the pre-mix tubes comprises an inlet adjacent the upstream face, an outlet adjacent the downstream face, and a channel extending between the inlet and the outlet. Each pre-mix tube also includes at least one fuel injector that at least partially extends outward from an exterior surface of the pre-mix tube, wherein the fuel injector is integrally formed with the pre-mix tube and is configured to facilitate fuel flow between the body portion and the channel.

  10. What happens inside a fuel cell? Developing an experimental functional map of fuel cell performance.

    Science.gov (United States)

    Brett, Daniel J L; Kucernak, Anthony R; Aguiar, Patricia; Atkins, Stephen C; Brandon, Nigel P; Clague, Ralph; Cohen, Lesley F; Hinds, Gareth; Kalyvas, Christos; Offer, Gregory J; Ladewig, Bradley; Maher, Robert; Marquis, Andrew; Shearing, Paul; Vasileiadis, Nikos; Vesovic, Velisa

    2010-09-10

    Fuel cell performance is determined by the complex interplay of mass transport, energy transfer and electrochemical processes. The convolution of these processes leads to spatial heterogeneity in the way that fuel cells perform, particularly due to reactant consumption, water management and the design of fluid-flow plates. It is therefore unlikely that any bulk measurement made on a fuel cell will accurately represent performance at all parts of the cell. The ability to make spatially resolved measurements in a fuel cell provides one of the most useful ways in which to monitor and optimise performance. This Minireview explores a range of in situ techniques being used to study fuel cells and describes the use of novel experimental techniques that the authors have used to develop an 'experimental functional map' of fuel cell performance. These techniques include the mapping of current density, electrochemical impedance, electrolyte conductivity, contact resistance and CO poisoning distribution within working PEFCs, as well as mapping the flow of reactant in gas channels using laser Doppler anemometry (LDA). For the high-temperature solid oxide fuel cell (SOFC), temperature mapping, reference electrode placement and the use of Raman spectroscopy are described along with methods to map the microstructural features of electrodes. The combination of these techniques, applied across a range of fuel cell operating conditions, allows a unique picture of the internal workings of fuel cells to be obtained and have been used to validate both numerical and analytical models.

  11. Fuel processors for fuel cell APU applications

    Science.gov (United States)

    Aicher, T.; Lenz, B.; Gschnell, F.; Groos, U.; Federici, F.; Caprile, L.; Parodi, L.

    The conversion of liquid hydrocarbons to a hydrogen rich product gas is a central process step in fuel processors for auxiliary power units (APUs) for vehicles of all kinds. The selection of the reforming process depends on the fuel and the type of the fuel cell. For vehicle power trains, liquid hydrocarbons like gasoline, kerosene, and diesel are utilized and, therefore, they will also be the fuel for the respective APU systems. The fuel cells commonly envisioned for mobile APU applications are molten carbonate fuel cells (MCFC), solid oxide fuel cells (SOFC), and proton exchange membrane fuel cells (PEMFC). Since high-temperature fuel cells, e.g. MCFCs or SOFCs, can be supplied with a feed gas that contains carbon monoxide (CO) their fuel processor does not require reactors for CO reduction and removal. For PEMFCs on the other hand, CO concentrations in the feed gas must not exceed 50 ppm, better 20 ppm, which requires additional reactors downstream of the reforming reactor. This paper gives an overview of the current state of the fuel processor development for APU applications and APU system developments. Furthermore, it will present the latest developments at Fraunhofer ISE regarding fuel processors for high-temperature fuel cell APU systems on board of ships and aircrafts.

  12. GSPEL - Fuel Cell Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — The Fuel Cell Lab (FCL) Provides testing for technology readiness of fuel cell systems The FCL investigates, tests and verifies the performance of fuel-cell systems...

  13. GSPEL - Fuel Cell Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — The Fuel Cell Lab (FCL)Provides testing for technology readiness of fuel cell systems The FCL investigates, tests and verifies the performance of fuel-cell systems...

  14. Fuel cells: A survey

    Science.gov (United States)

    Crowe, B. J.

    1973-01-01

    A survey of fuel cell technology and applications is presented. The operating principles, performance capabilities, and limitations of fuel cells are discussed. Diagrams of fuel cell construction and operating characteristics are provided. Photographs of typical installations are included.

  15. Future aviation fuels overview

    Science.gov (United States)

    Reck, G. M.

    1980-01-01

    The outlook for aviation fuels through the turn of the century is briefly discussed and the general objectives of the NASA Lewis Alternative Aviation Fuels Research Project are outlined. The NASA program involves the evaluation of potential characteristics of future jet aircraft fuels, the determination of the effects of those fuels on engine and fuel system components, and the development of a component technology to use those fuels.

  16. Catalytic Fuel Conversion Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This facility enables unique catalysis research related to power and energy applications using military jet fuels and alternative fuels. It is equipped with research...

  17. Fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Enomoto, Hirofumi.

    1989-05-22

    This invention aims to maintain a long-term operation with stable cell output characteristics by uniformly supplying an electrolyte from the reserver to the matrix layer over the entire matrix layer, and further to prevent the excessive wetting of the catalyst layer by smoothly absorbing the volume change of the electrolyte, caused by the repeated stop/start-up of the fuel cell, within the reserver system. For this purpose, in this invention, an electrolyte transport layer, which connects with an electrolyte reservor formed at the electrode end, is partly formed between the electrode material and the catalyst layer; a catalyst layer, which faces the electrolyte transport layer, has through-holes, which connect to the matrix, dispersely distributed. The electrolyte-transport layer is a thin sheet of a hydrophilic fibers which are non-wovens of such fibers as carbon, silicon carbide, silicon nitride or inorganic oxides. 11 figs.

  18. Multi-Channel Retailing

    Directory of Open Access Journals (Sweden)

    Dirk Morschett, Dr.,

    2005-01-01

    Full Text Available Multi-channel retailing entails the parallel use by retailing enterprises of several sales channels. The results of an online buyer survey which has been conducted to investigate the impact of multi-channel retailing (i.e. the use of several retail channels by one retail company on consumer behaviour show that the frequently expressed concern that the application of multi-channel systems in retailing would be associated with cannibalization effects, has proven unfounded. Indeed, the appropriate degree of similarity, consistency, integration and agreement achieves the exact opposite. Different channels create different advantages for consumers. Therefore the total benefit an enterprise which has a multi-channel system can offer to its consumers is larger, the greater the number of available channels. The use of multi-channel systems is associated with additional purchases in the different channels. Such systems are thus superior to those offering only one sales channel to their customers. Furthermore, multi-channel systems with integrated channels are superior to those in which the channels are essentially autonomous and independent of one another. In integrated systems, consumers can achieve synergy effects in the use of sales-channel systems. Accordingly, when appropriately formulated, multi-channel systems in retailing impact positively on consumers. They use the channels more frequently, buy more from them and there is a positive customer-loyalty impact. Multi-channel systems are strategic options for achieving customer loyalty, exploiting customer potential and for winning new customers. They are thus well suited for approaching differing and varied target groups.

  19. Annual report 1998-1999

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-07-01

    This is the Annual Report of the Atomic Energy of Canada Limited for the year ending March 31, 1999 and summarizes the activities of AECL during the period 1998-1999. The Activities covered in this Report include the CANDU Reactor Business, with excellent progress reported on the construction of two 700 MWe-class CANDU reactors in Qinshan, China. In the Republic of Korea, Wolsong Unit entered into commercial operation and Wolsong Unit 4 achieved sustained nuclear reaction. The Report also covers AECL's R and D and Waste Management programs. In the R and D section, the report outlines the development of the CANFLEX fuel bundle, Fuel Channels, Reactor Safety, Code Validation, Fuels and Fuel Cycles as well as Heavy Water production. Progress in the Waste Management program is also discussed.

  20. Solid polymer MEMS-based fuel cells

    Science.gov (United States)

    Jankowski, Alan F.; Morse, Jeffrey D.

    2008-04-22

    A micro-electro-mechanical systems (MEMS) based thin-film fuel cells for electrical power applications. The MEMS-based fuel cell may be of a solid oxide type (SOFC), a solid polymer type (SPFC), or a proton exchange membrane type (PEMFC), and each fuel cell basically consists of an anode and a cathode separated by an electrolyte layer. The electrolyte layer can consist of either a solid oxide or solid polymer material, or proton exchange membrane electrolyte materials may be used. Additionally catalyst layers can also separate the electrodes (cathode and anode) from the electrolyte. Gas manifolds are utilized to transport the fuel and oxidant to each cell and provide a path for exhaust gases. The electrical current generated from each cell is drawn away with an interconnect and support structure integrated with the gas manifold. The fuel cells utilize integrated resistive heaters for efficient heating of the materials. By combining MEMS technology with thin-film deposition technology, thin-film fuel cells having microflow channels and full-integrated circuitry can be produced that will lower the operating temperature an will yield an order of magnitude greater power density than the currently known fuel cells.

  1. Solid oxide MEMS-based fuel cells

    Science.gov (United States)

    Jankowksi, Alan F.; Morse, Jeffrey D.

    2007-03-13

    A micro-electro-mechanical systems (MEMS) based thin-film fuel cells for electrical power applications. The MEMS-based fuel cell may be of a solid oxide type (SOFC), a solid polymer type (SPFC), or a proton exchange membrane type (PEMFC), and each fuel cell basically consists of an anode and a cathode separated by an electrolyte layer. The electrolyte layer can consist of either a solid oxide or solid polymer material, or proton exchange membrane electrolyte materials may be used. Additionally catalyst layers can also separate the electrodes (cathode and anode) from the electrolyte. Gas manifolds are utilized to transport the fuel and oxidant to each cell and provide a path for exhaust gases. The electrical current generated from each cell is drawn away with an interconnect and support structure integrated with the gas manifold. The fuel cells utilize integrated resistive heaters for efficient heating of the materials. By combining MEMS technology with thin-film deposition technology, thin-film fuel cells having microflow channels and full-integrated circuitry can be produced that will lower the operating temperature an will yield an order of magnitude greater power density than the currently known fuel cells.

  2. Development of nuclear fuel for integrated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kee Nam; Kim, H. K.; Kang, H. S.; Yoon, K. H.; Chun, T. H.; In, W. K.; Oh, D. S.; Kim, D. W.; Woo, Y. M

    1999-04-01

    The spacer grid assembly which provides both lateral and vertical support for the fuel rods and also provides a flow channel between the fuel rods to afford the heat transfer from the fuel pellet into the coolant in a reactor, is one of the major structural components of nuclear fuel for LWR. Therefore, the spacer grid assembly is a highly ranked component when the improvement of hardware is pursued for promoting fuel performance. Main objective of this project is to develop the inherent spacer grid assembly and to research relevant technologies on the spacer grid assembly. And, the UO{sub 2}-based SMART fuel is preliminarily designed for the 330MWt class SMART, which is planned to produce heat as well as electricity. Results from this project are listed as follows. 1. Three kinds of spacer grid candidates have been invented and applied for domestic and US patents. In addition, the demo SG(3x3 array) were fabricated, which the mechanical/structural test was carried out with. 2. The mechanical/structural technologies related to the spacer grid development are studied and relevant test requirements were established. 3. Preliminary design data of the UO{sub 2}-based SMART fuel have been produced. The structural characteristics of several components such as the top/bottom end piece and the holddown spring assembly were analysed by consulting the numerical method.

  3. USACE Navigation Channels 2012

    Data.gov (United States)

    California Department of Resources — This dataset represents both San Francisco and Los Angeles District navigation channel lines. All San Francisco District channel lines were digitized from CAD files...

  4. Development of laser welded appendages to Zircaloy-4 fuel tubing (sheath/cladding)

    Energy Technology Data Exchange (ETDEWEB)

    Livingstone, S., E-mail: steve.livingstone@cnl.ca [Canadian Nuclear Laboratories Limited, Chalk River, ON, Canada K0J 1J0 (Canada); Xiao, L. [Canadian Nuclear Laboratories Limited, Chalk River, ON, Canada K0J 1J0 (Canada); Corcoran, E.C.; Ferrier, G.A.; Potter, K.N. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, Kingston, ON, Canada K7K 7B4 (Canada)

    2015-04-01

    Highlights: • Examines feasibility of laser welding appendages to Zr-4 tubing. • Laser welding minimizes the HAZ and removes toxic Be. • Mechanical properties of laser welds appear competitive with induction brazed joints. • Work appears promising and lays the foundation for further investigations. - Abstract: Laser welding is a potential alternative to the induction brazing process commonly used for appendage attachment in CANDU{sup ®} fuel fabrication that uses toxic Be as a filler metal, and creates multiple large heat affected zones in the sheath. For this work, several appendages were laser welded to tubing using different laser heat input settings and then examined with a variety of techniques: visual examination, metallography, shear strength testing, impact testing, and fracture surface analysis. Where possible, the examination results are contrasted against production induction brazed joints. The work to date looks promising for laser welded appendages. Further work on joint optimization, corrosion testing, irradiation testing, and post-irradiation examination will be performed in the future.

  5. HTGR fuel and fuel cycle technology

    Energy Technology Data Exchange (ETDEWEB)

    Lotts, A.L.; Coobs, J.H.

    1976-08-01

    The status of fuel and fuel cycle technology for high-temperature gas-cooled reactors (HTGRs) is reviewed. The all-ceramic core of the HTGRs permits high temperatures compared with other reactors. Core outlet temperatures of 740/sup 0/C are now available for the steam cycle. For advanced HTGRs such as are required for direct-cycle power generation and for high-temperature process heat, coolant temperatures as high as 1000/sup 0/C may be expected. The paper discusses the variations of HTGR fuel designs that meet the performance requirements and the requirements of the isotopes to be used in the fuel cycle. Also discussed are the fuel cycle possibilities, which include the low-enrichment cycle, the Th-/sup 233/U cycle, and plutonium utilization in either cycle. The status of fuel and fuel cycle development is summarized.

  6. Carbon fuel cells with carbon corrosion suppression

    Science.gov (United States)

    Cooper, John F [Oakland, CA

    2012-04-10

    An electrochemical cell apparatus that can operate as either a fuel cell or a battery includes a cathode compartment, an anode compartment operatively connected to the cathode compartment, and a carbon fuel cell section connected to the anode compartment and the cathode compartment. An effusion plate is operatively positioned adjacent the anode compartment or the cathode compartment. The effusion plate allows passage of carbon dioxide. Carbon dioxide exhaust channels are operatively positioned in the electrochemical cell to direct the carbon dioxide from the electrochemical cell.

  7. Recent Developments in the Management of Cameco Corporation's Fuel Services Division Waste - 13144

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Thomas P. [Cameco Corporation, Port Hope, Ontario (Canada)

    2013-07-01

    Cameco Corporation is a world leader in uranium production. Headquartered in Saskatoon, Saskatchewan our operations provide 16% of the world uranium mine production and we have approximately 435 million pounds of proven and probable uranium reserves. Cameco mining operations are located in Saskatchewan, Wyoming, Nebraska and Kazakhstan. Cameco is also a major supplier of uranium processing services required to produce fuel for the generation of clean energy. These operations are based in Blind River, Cobourg and Port Hope, Ontario and are collectively referred to as the Fuel Services Division. The Fuel Services Division produces uranium trioxide from uranium ore concentrate at the Blind River Refinery. Cameco produces uranium hexafluoride and uranium dioxide at the Port Hope Conversion Facility. Cameco operates a fuel manufacturing facility in Port Hope, Ontario and a metal fabrication facility located in Cobourg, Ontario. The company manufactures fuel bundles utilized in the Candu reactors. Cameco's Fuel Services Division produces several types of low-level radioactively contaminated wastes. Internal processing capabilities at both the Blind River Refinery and Port Hope Conversion Facility are extensive and allow for the recycling of several types of waste. Notwithstanding these capabilities there are certain wastes that are not amenable to the internal processing capabilities and must be disposed of appropriately. Disposal options for low-level radioactively contaminated wastes in Canada are limited primarily due to cost considerations. In recent years, Cameco has started to ship marginally contaminated wastes (<500 ppm uranium) to the United States for disposal in an appropriate landfill. The landfill is owned by US Ecology Incorporated and is located near Grand View, Idaho 70 miles southeast of Boise in the Owyhee Desert. The facility treats and disposes hazardous waste, non-hazardous industrial waste and low-activity radioactive material. The site

  8. KV7 potassium channels

    DEFF Research Database (Denmark)

    Stott, Jennifer B; Jepps, Thomas Andrew; Greenwood, Iain A

    2014-01-01

    Potassium channels are key regulators of smooth muscle tone, with increases in activity resulting in hyperpolarisation of the cell membrane, which acts to oppose vasoconstriction. Several potassium channels exist within smooth muscle, but the KV7 family of voltage-gated potassium channels have been...

  9. Quantum Multiple Access Channel

    Institute of Scientific and Technical Information of China (English)

    侯广; 黄民信; 张永德

    2002-01-01

    We consider the transmission of classical information over a quantum channel by many senders, which is a generalization of the two-sender case. The channel capacity region is shown to be a convex hull bound by the yon Neumann entropy and the conditional yon Neumann entropies. The result allows a reasonable distribution of channel capacity over the senders.

  10. Cardiac potassium channel subtypes

    DEFF Research Database (Denmark)

    Schmitt, Nicole; Grunnet, Morten; Olesen, Søren-Peter

    2014-01-01

    About 10 distinct potassium channels in the heart are involved in shaping the action potential. Some of the K(+) channels are primarily responsible for early repolarization, whereas others drive late repolarization and still others are open throughout the cardiac cycle. Three main K(+) channels...

  11. Alternative Fuels (Briefing Charts)

    Science.gov (United States)

    2009-06-19

    SASOL Jet Fuel Qatar GTL Syntroleum Jet fuel in B-52 Nigeria GTL China Coal GTL Shell Bintulu GTL Cellulose ethanol for ground use Ocean Bio...Bintulu GTL Ocean Bio-fuel Factories Bio-butanol for ground use Future Energy Source Resurgence in Nuclear Power Bio-jet tests done ASTM Spec...High energy deoxygenated bio-jet fuel from algae PSU coal derived JP-8 B-52 emissions Scoping study HBR TF emissions New bio- fuel impacts Adv

  12. Multiphase transport in polymer electrolyte membrane fuel cells

    Science.gov (United States)

    Gauthier, Eric D.

    Polymer electrolyte membrane fuel cells (PEMFCs) enable efficient conversion of fuels to electricity. They have enormous potential due to the high energy density of the fuels they utilize (hydrogen or alcohols). Power density is a major limitation to wide-scale introduction of PEMFCs. Power density in hydrogen fuel cells is limited by accumulation of water in what is termed fuel cell `flooding.' Flooding may occur in either the gas diffusion layer (GDL) or within the flow channels of the bipolar plate. These components comprise the electrodes of the fuel cell and balance transport of reactants/products with electrical conductivity. This thesis explores the role of electrode materials in the fuel cell and examines the fundamental connection between material properties and multiphase transport processes. Water is generated at the cathode catalyst layer. As liquid water accumulates it will utilize the largest pores in the GDL to go from the catalyst layer to the flow channels. Water collects to large pores via lateral transport at the interface between the GDL and catalyst layer. We have shown that water may be collected in these large pores from several centimeters away, suggesting that we could engineer the GDL to control flooding with careful placement and distribution of large flow-directing pores. Once liquid water is in the flow channels it forms slugs that block gas flow. The slugs are pushed along the channel by a pressure gradient that is dependent on the material wettability. The permeable nature of the GDL also plays a major role in slug growth and allowing bypass of gas between adjacent channels. Direct methanol fuel cells (DMFCs) have analogous multiphase flow issues where carbon dioxide bubbles accumulate, `blinding' regions of the fuel cell. This problem is fundamentally similar to water management in hydrogen fuel cells but with a gas/liquid phase inversion. Gas bubbles move laterally through the porous GDL and emerge to form large bubbles within the

  13. Mathematical Modeling Analysis and Optimization of Key Design Parameters of Proton-Conductive Solid Oxide Fuel Cells

    OpenAIRE

    Liu, Hong; Akhtar, Zoheb; Li, Peiwen; Wang, Kai

    2014-01-01

    A proton-conductive solid oxide fuel cell (H-SOFC) has the advantage of operating at higher temperatures than a PEM fuel cell, but at lower temperatures than a SOFC. This study proposes a mathematical model for an H-SOFC in order to simulate the performance and optimize the flow channel designs. The model analyzes the average mass transfer and species’ concentrations in flow channels, which allows the determination of an average concentration polarization in anode and cathode gas channels, th...

  14. Mathematical Modeling Analysis and Optimization of Key Design Parameters of Proton-Conductive Solid Oxide Fuel Cells

    OpenAIRE

    Hong Liu; Zoheb Akhtar; Peiwen Li; Kai Wang

    2014-01-01

    A proton-conductive solid oxide fuel cell (H-SOFC) has the advantage of operating at higher temperatures than a PEM fuel cell, but at lower temperatures than a SOFC. This study proposes a mathematical model for an H-SOFC in order to simulate the performance and optimize the flow channel designs. The model analyzes the average mass transfer and species’ concentrations in flow channels, which allows the determination of an average concentration polarization in anode and cathode gas channels, t...

  15. Protocol channels as a new design alternative of covert channels

    CERN Document Server

    Wendzel, Steffen

    2008-01-01

    Covert channel techniques are used by attackers to transfer hidden data. There are two main categories of covert channels: timing channels and storage channels. This paper introduces a third category called protocol channels. A protocol channel switches one of at least two protocols to send a bit combination to a destination while sent packets include no hidden information themselves.

  16. Surface vacancy channels through ion channeling

    Energy Technology Data Exchange (ETDEWEB)

    Redinger, Alex; Standop, Sebastian; Michely, Thomas [II. Physikalisches Institut, Universitaet Koeln, Zuelpicher Strasse 77, 50937 Koeln (Germany); Rosandi, Yudi; Urbassek, Herbert M. [Fachbereich Physik, Technische Universitaet Kaiserslautern, Erwin-Schroedinger-Strasse, D-67663 Kaiserslautern (Germany)

    2009-07-01

    Damage patterns of single ion impacts on Pt(111) have been studied by scanning tunneling microscopy (STM) and molecular dynamics simulations (MD). Low temperature experiments, where surface diffusion is absent, have been performed for argon and xenon ions with energies between 1 keV and 15 keV at an angle of incidence of 86 {sup circle} measured with respect to the surface normal. Ions hitting preexisting illuminated step edges penetrate into the crystal and are guided in open crystallographic directions, one or more layers underneath the surface (subsurface channeling). In the case of argon channeling the resulting surface damage consists of adatom and vacancy pairs aligned in ion beam direction. After xenon channeling thin surface vacancy trenches along the ion trajectories - surface vacancy channels - are observed. They result from very efficient sputtering and adatom production along the ion trajectory. This phenomena is well reproduced in molecular dynamics simulations of single ion impacts at 0 K. The damage patterns of Argon and Xenon impacts can be traced back to the different energy losses of the particles in the channel. Channeling distances exceeding 1000 A for 15 keV xenon impacts are observed.

  17. Capillary-Driven Flow in Liquid Filaments Connecting Orthogonal Channels

    Science.gov (United States)

    Allen, Jeffrey S.

    2005-01-01

    Capillary phenomena plays an important role in the management of product water in PEM fuel cells because of the length scales associated with the porous layers and the gas flow channels. The distribution of liquid water within the network of gas flow channels can be dramatically altered by capillary flow. We experimentally demonstrate the rapid movement of significant volumes of liquid via capillarity through thin liquid films which connect orthogonal channels. The microfluidic experiments discussed provide a good benchmark against which the proper modeling of capillarity by computational models may be tested. The effect of surface wettability, as expressed through the contact angle, on capillary flow will also be discussed.

  18. HTGR Fuel performance basis

    Energy Technology Data Exchange (ETDEWEB)

    Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

    1982-05-01

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600/sup 0/C, and complete fuel failure occurs at 2660/sup 0/C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents.

  19. Deployment of advanced MACSTOR dry spent fuel storage technology in Korea - A joint development program

    Energy Technology Data Exchange (ETDEWEB)

    Cobanoglu, M. M.; Pattantyus, P. [Atomic Energy Canada Limited, Ottawa (Canada); Song, M. J.; Lee, H. Y. [KHNP/NETEC, Daejeon (Korea, Republic of)

    2002-04-15

    KHNP/NETEC's (K/N) and Atomic Energy of Canada Limited (AECL) are undertaking to jointly develop a high capacity dry storage structure made of reinforced concrete that uses the MACSTOR storage module concept. This effort is based on AECL's experience and on the successful deployment of concrete canisters at Wolsong and on the deployment of air-cooled MACSTOR modules at the Gentilly 2 reactor in Canada. The proposed approach addresses the conditions specific to the Wolsong site: large yearly fuel throughput, space limitations and the need for an economical dry storage structure that can store lifetime spent fuel inventories expected from the four CANDU units. The selected configuration is a 4-row MACSTOR module with a capacity of 24,000 bundles stored in 400 baskets, each holding 60 spent fuel bundles. The module is thus termed MACSTOR/KN-400 and is expected to offer a repetitive storage density increase by a factor of approximately 3, compared to concrete canisters presently used. The four Wolsong units generate spent fuel bundles that, with the high capacity factors achieved, are in the order of 20,000 bundles or more per year. At all Korean nuclear facilities, space limitations dictate the need for storage structures having high storage density. Storage density increases have to be accomplished while maintaining safety parameters during the full term storage of nuclear fuel. During the early 1990's AECL has proceeded with the development of a 2-row MACSTOR storage module that offered a higher storage density and a more economical solution compared to the stand alone concrete canister used at Wolsong 1. These modules are in use at Gentilly since the mid 1990's and operate at a capacity of 200 baskets. The selection of a MACSTOR module with 4 rows of storage cylinders is the natural evolution of the already deployed configuration. It can be developed without additional thermal testing as the fuel is maintained within the existing licensing

  20. A sensitivity study on neutronic properties of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The {sup 239}Pu and {sup 235}U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the feed uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel and shown that it is desirable to increase the {sup 239}Pu and {sup 235}U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking factor, which is a measure of the reactor trip margin, increases with the total fissile content. Considering these neutronic characteristics of the DUPIC fuel, it is recommended to have enrichments of 0.45 and 1.00 wt% for {sup 239}Pu and {sup 235}U, respectively. 3 refs., 2 tabs. (Author)

  1. Advances in PEM fuel cells with CFD techniques

    Energy Technology Data Exchange (ETDEWEB)

    Robalinho, Eric; Cunha, Edgar Ferrari da; Zararya, Ahmed; Linardi, Marcelo [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)], Email: eric@ipen.br; Cekinski, Efrain [Instituto de Pesquisas Tecnologicas (IPT), Sao Paulo, SP (Brazil)

    2010-07-01

    This paper presents some applications of computational fluid dynamics techniques in the optimization of Proton Exchange Membrane Fuel Cell (PEMFC) designs. The results concern: modeling of gas distribution channels, the study for both porous anode and cathode and the three-dimensional modeling of a partial geometry layer containing catalytic Gas Diffusion Layers (GDL) and membrane. Numerical results of the simulations of graphite plates flow channels, using ethanol as fuel, are also presented. Some experimental results are compared to the corresponding numerical ones for several cases, demonstrating the importance and usefulness of this computational tool. (author)

  2. Materials for fuel cells

    Directory of Open Access Journals (Sweden)

    Sossina M Haile

    2003-03-01

    Full Text Available Because of their potential to reduce the environmental impact and geopolitical consequences of the use of fossil fuels, fuel cells have emerged as tantalizing alternatives to combustion engines. Like a combustion engine, a fuel cell uses some sort of chemical fuel as its energy source but, like a battery, the chemical energy is directly converted to electrical energy, without an often messy and relatively inefficient combustion step. In addition to high efficiency and low emissions, fuel cells are attractive for their modular and distributed nature, and zero noise pollution. They will also play an essential role in any future hydrogen fuel economy.

  3. A review of qualitative inspection aspects of end fittings in an Indian pressurized heavy water reactor

    OpenAIRE

    Urva Pancholi; Dhaval Dave; Ajay Patel

    2016-01-01

    The paper provides a summarized description of the current state of knowledge and practices used in India, in the qualitative inspection of end fittings – a key component of the fuel channel assembly of a pressurized heavy water reactor (PHWR), generally of a Canadian Deuterium Uranium (CANDU) type. Further it discusses various quality inspection techniques; and the high standards and mechanical precision of the job required, to be accepted as viable nuclear reactor component. The techniqu...

  4. Flow rate and humidification effects on a PEM fuel cell performance and operation

    Science.gov (United States)

    Guvelioglu, Galip H.; Stenger, Harvey G.

    A new algorithm is presented to integrate component balances along polymer electrolyte membrane fuel cell (PEMFC) channels to obtain three-dimensional results from a detailed two-dimensional finite element model. The analysis studies the cell performance at various hydrogen flow rates, air flow rates and humidification levels. This analysis shows that hydrogen and air flow rates and their relative humidity are critical to current density, membrane dry-out, and electrode flooding. Uniform current densities along the channels are known to be critical for thermal management and fuel cell life. This approach, of integrating a detailed two-dimensional across-the-channel model, is a promising method for fuel cell design due to its low computational cost compared to three-dimensional computational fluid dynamics models, its applicability to a wide range of fuel cell designs, and its ease of extending to fuel cell stack models.

  5. Quantum broadcast channels

    CERN Document Server

    Yard, J; Devetak, I; Yard, Jon; Hayden, Patrick; Devetak, Igor

    2006-01-01

    We analyze quantum broadcast channels, which are quantum channels with a single sender and many receivers. Focusing on channels with two receivers for simplicity, we generalize a number of results from the network Shannon theory literature which give the rates at which two senders can receive a common message, while a personalized one is sent to one of them. Our first collection of results applies to channels with a classical input and quantum outputs. The second class of theorems we prove concern sending a common classical message over a quantum broadcast channel, while sending quantum information to one of the receivers. The third group of results we obtain concern communication over an isometry, giving the rates at quantum information can be sent to one receiver, while common quantum information is sent to both, in the sense that tripartite GHZ entanglement is established. For each scenario, we provide an additivity proof for an appropriate class of channels, yielding single-letter characterizations of the...

  6. Quantum feedback channels

    CERN Document Server

    Bowen, G

    2002-01-01

    In classical information theory the capacity of a noisy communication channel cannot be increased by the use of feedback. In quantum information theory the no-cloning theorem means that noiseless copying and feedback of quantum information cannot be achieved. In this paper, quantum feedback is defined as the unlimited use of a noiseless quantum channel from receiver to sender. Given such quantum feedback, it is shown to provide no increase in the entanglement-assisted capacities of a noisy quantum channel, in direct analogy to the classical case. It is also shown that in various cases of non-assisted capacities, feedback can increase the capacity of many quantum channels.

  7. Ion channels in asthma.

    Science.gov (United States)

    Valverde, Miguel A; Cantero-Recasens, Gerard; Garcia-Elias, Anna; Jung, Carole; Carreras-Sureda, Amado; Vicente, Rubén

    2011-09-23

    Ion channels are specialized transmembrane proteins that permit the passive flow of ions following their electrochemical gradients. In the airways, ion channels participate in the production of epithelium-based hydroelectrolytic secretions and in the control of intracellular Ca(2+) levels that will ultimately activate almost all lung cells, either resident or circulating. Thus, ion channels have been the center of many studies aiming to understand asthma pathophysiological mechanisms or to identify therapeutic targets for better control of the disease. In this minireview, we focus on molecular, genetic, and animal model studies associating ion channels with asthma.

  8. TRP channels in schistosomes

    Directory of Open Access Journals (Sweden)

    Swarna Bais

    2016-12-01

    Full Text Available Praziquantel (PZQ is effectively the only drug currently available for treatment and control of schistosomiasis, a disease affecting hundreds of millions of people worldwide. Many anthelmintics, likely including PZQ, target ion channels, membrane protein complexes essential for normal functioning of the neuromusculature and other tissues. Despite this fact, only a few classes of parasitic helminth ion channels have been assessed for their pharmacological properties or for their roles in parasite physiology. One such overlooked group of ion channels is the transient receptor potential (TRP channel superfamily. TRP channels share a common core structure, but are widely diverse in their activation mechanisms and ion selectivity. They are critical to transducing sensory signals, responding to a wide range of external stimuli. They are also involved in other functions, such as regulating intracellular calcium and organellar ion homeostasis and trafficking. Here, we review current literature on parasitic helminth TRP channels, focusing on those in schistosomes. We discuss the likely roles of these channels in sensory and locomotor activity, including the possible significance of a class of TRP channels (TRPV that is absent in schistosomes. We also focus on evidence indicating that at least one schistosome TRP channel (SmTRPA has atypical, TRPV1-like pharmacological sensitivities that could potentially be exploited for future therapeutic targeting.

  9. FUEL CELL ELECTRODE MATERIALS

    Science.gov (United States)

    FUEL CELL ELECTRODE MATERIALS. RAW MATERIAL SELECTION INFLUENCES POLARIZATION BUT IS NOT A SINGLE CONTROLLING FACTOR. AVAILABLE...DATA INDICATES THAT AN INTERRELATIONSHIP OF POROSITY, AVERAGE PORE VOLUME, AND PERMEABILITY CONTRIBUTES TO ELECTRODE FUEL CELL BEHAVIOR.

  10. Navy Fuel Specification Standardization

    Science.gov (United States)

    1992-04-01

    surfaced periodically to convert further to a single-fuel operation, i.e., one fuel for both aircraft and ship propulsion /power systems. This study...lead to the development of a single distillate fuel for ship propulsion , resulting eventually in the MIL-F-16884 Naval Distillate Fuel (NDF) used today...for both aircraft and ship propulsion /power systems. This report summarizes a study to consider this problem in light of current systems and

  11. Direct hydrocarbon fuel cells

    Science.gov (United States)

    Barnett, Scott A.; Lai, Tammy; Liu, Jiang

    2010-05-04

    The direct electrochemical oxidation of hydrocarbons in solid oxide fuel cells, to generate greater power densities at lower temperatures without carbon deposition. The performance obtained is comparable to that of fuel cells used for hydrogen, and is achieved by using novel anode composites at low operating temperatures. Such solid oxide fuel cells, regardless of fuel source or operation, can be configured advantageously using the structural geometries of this invention.

  12. A linearization of quantum channels

    Science.gov (United States)

    Crowder, Tanner

    2015-06-01

    Because the quantum channels form a compact, convex set, we can express any quantum channel as a convex combination of extremal channels. We give a Euclidean representation for the channels whose inverses are also valid channels; these are a subset of the extreme points. They form a compact, connected Lie group, and we calculate its Lie algebra. Lastly, we calculate a maximal torus for the group and provide a constructive approach to decomposing any invertible channel into a product of elementary channels.

  13. Alternative Fuels Data Center

    Energy Technology Data Exchange (ETDEWEB)

    None

    2013-06-01

    Fact sheet describes the Alternative Fuels Data Center, which provides information, data, and tools to help fleets and other transportation decision makers find ways to reduce petroleum consumption through the use of alternative and renewable fuels, advanced vehicles, and other fuel-saving measures.

  14. Modeling: driving fuel cells

    Directory of Open Access Journals (Sweden)

    Michael Francis

    2002-05-01

    Fuel cells were invented in 1839 by Sir William Grove, a Welsh judge and gentleman scientist, as a result of his experiments on the electrolysis of water. To put it simply, fuel cells are electrochemical devices that take hydrogen gas from fuel, combine it with oxygen from the air, and generate electricity and heat, with water as the only by-product.

  15. Ion channels in toxicology.

    Science.gov (United States)

    Restrepo-Angulo, Iván; De Vizcaya-Ruiz, Andrea; Camacho, Javier

    2010-08-01

    Ion channels play essential roles in human physiology and toxicology. Cardiac contraction, neural transmission, temperature sensing, insulin release, regulation of apoptosis, cellular pH and oxidative stress, as well as detection of active compounds from chilli, are some of the processes in which ion channels have an important role. Regulation of ion channels by several chemicals including those found in air, water and soil represents an interesting potential link between environmental pollution and human diseases; for instance, de novo expression of ion channels in response to exposure to carcinogens is being considered as a potential tool for cancer diagnosis and therapy. Non-specific binding of several drugs to ion channels is responsible for a huge number of undesirable side-effects, and testing guidelines for several drugs now require ion channel screening for pharmaceutical safety. Animal toxins targeting human ion channels have serious effects on the population and have also provided a remarkable tool to study the molecular structure and function of ion channels. In this review, we will summarize the participation of ion channels in biological processes extensively used in toxicological studies, including cardiac function, apoptosis and cell proliferation. Major findings on the adverse effects of drugs on ion channels as well as the regulation of these proteins by different chemicals, including some pesticides, are also reviewed. Association of ion channels and toxicology in several biological processes strongly suggests these proteins to be excellent candidates to follow the toxic effects of xenobiotics, and as potential early indicators of life-threatening situations including chronic degenerative diseases.

  16. Fuel-flexible burner apparatus and method for fired heaters

    Energy Technology Data Exchange (ETDEWEB)

    Zink, Darton J.; Isaacs, Rex K.; Jamaluddin, A. S. (Jamal); Benson, Charles E.; Pellizzari, Roberto O.; Little, Cody L.; Marty, Seth A.; Imel, K. Parker; Barnes, Jonathon E.; Parker, Chris S.

    2017-03-14

    A burner apparatus for a fired heating system and a method of burner operation. The burner provides stable operation when burning gas fuels having heating values ranging from low to high and accommodates sudden wide changes in the Wobbe value of the fuel delivered to the burner. The burner apparatus includes a plurality of exterior fuel ejectors and has an exterior notch which extends around the burner wall for receiving and combusting a portion of the gas fuel. At least a portion of the hot combustion product gas produced in the exterior notch is delivered through channels formed in the burner wall to the combustion area at the forward end of the burner. As the Wobbe value of the gas fuel decreases, one or more outer series of addition ejectors can be automatically activated as needed to maintain the amount of heat output desired.

  17. What Happens Inside a Fuel Cell? Developing an Experimental Functional Map of Fuel Cell Performance

    KAUST Repository

    Brett, Daniel J. L.

    2010-08-20

    Fuel cell performance is determined by the complex interplay of mass transport, energy transfer and electrochemical processes. The convolution of these processes leads to spatial heterogeneity in the way that fuel cells perform, particularly due to reactant consumption, water management and the design of fluid-flow plates. It is therefore unlikely that any bulk measurement made on a fuel cell will accurately represent performance at all parts of the cell. The ability to make spatially resolved measurements in a fuel cell provides one of the most useful ways in which to monitor and optimise performance. This Minireview explores a range of in situ techniques being used to study fuel cells and describes the use of novel experimental techniques that the authors have used to develop an \\'experimental functional map\\' of fuel cell performance. These techniques include the mapping of current density, electrochemical impedance, electrolyte conductivity, contact resistance and CO poisoning distribution within working PEFCs, as well as mapping the flow of reactant in gas channels using laser Doppler anemometry (LDA). For the high-temperature solid oxide fuel cell (SOFC), temperature mapping, reference electrode placement and the use of Raman spectroscopy are described along with methods to map the microstructural features of electrodes. The combination of these techniques, applied across a range of fuel cell operating conditions, allows a unique picture of the internal workings of fuel cells to be obtained and have been used to validate both numerical and analytical models. © 2010 Wiley-VCH Verlag GmbH& Co. KGaA, Weinheim.

  18. Athermalized channeled spectropolarimeter enhancement.

    Energy Technology Data Exchange (ETDEWEB)

    Jones, Julia Craven; Way, Brandyn Michael; Mercier, Jeffrey Alan; Hunt, Jeffery P.

    2013-09-01

    Channeled spectropolarimetry can measure the complete polarization state of light as a function of wavelength. Typically, a channeled spectropolarimeter uses high order retarders made of uniaxial crystal to amplitude modulate the measured spectrum with the spectrally-dependent Stokes polarization information. A primary limitation of conventional channeled spectropolarimeters is related to the thermal variability of the retarders. Thermal variation often forces frequent system recalibration, particularly for field deployed systems. However, implementing thermally stable retarders, made of biaxial crystal, results in an athermal channeled spectropolarimeter that relieves the need for frequent recalibration. This report presents experimental results for an anthermalized channeled spectropolarimeter prototype produced using potassium titanyl phosphate. The results of this prototype are compared to the current thermal stabilization state of the art. Finally, the application of the technique to the thermal infrared is studied, and the athermalization concept is applied to an infrared imaging spectropolarimeter design.

  19. Alternative aviation turbine fuels

    Science.gov (United States)

    Grobman, J.

    1977-01-01

    The efficient utilization of fossil fuels by future jet aircraft may necessitate the broadening of current aviation turbine fuel specifications. The most significant changes in specifications would be an increased aromatics content and a higher final boiling point in order to minimize refinery energy consumption and costs. These changes would increase the freezing point and might lower the thermal stability of the fuel and could cause increased pollutant emissions, increased smoke and carbon formation, increased combustor liner temperatures, and poorer ignition characteristics. This paper discusses the effects that broadened specification fuels may have on present-day jet aircraft and engine components and the technology required to use fuels with broadened specifications.

  20. Fuel cells : a viable fossil fuel alternative

    Energy Technology Data Exchange (ETDEWEB)

    Paduada, M.

    2007-02-15

    This article presented a program initiated by Natural Resources Canada (NRCan) to develop proof-of-concept of underground mining vehicles powered by fuel cells in order to eliminate emissions. Recent studies on American and Canadian underground mines provided the basis for estimating the operational cost savings of switching from diesel to fuel cells. For the Canadian mines evaluated, the estimated ventilation system operating cost reductions ranged from 29 per cent to 75 per cent. In order to demonstrate the viability of a fuel cell-powered vehicle, NRCan has designed a modified Caterpillar R1300 loader with a 160 kW hybrid power plant in which 3 stacks of fuel cells deliver up to 90 kW continuously, and a nickel-metal hydride battery provides up to 70 kW. The battery subsystem transiently boosts output to meet peak power requirements and also accommodates regenerative braking. Traction for the loader is provided by a brushless permanent magnet traction motor. The hydraulic pump motor is capable of a 55 kW load continuously. The loader's hydraulic and traction systems are operated independently. Future fuel cell-powered vehicles designed by the program may include a locomotive and a utility vehicle. Future mines running their operations with hydrogen-fueled equipment may also gain advantages by employing fuel cells in the operation of handheld equipment such as radios, flashlights, and headlamps. However, the proton exchange membrane (PEM) fuel cells used in the project are prohibitively expensive. The catalytic content of a fuel cell can add hundreds of dollars per kW of electric output. Production of catalytic precious metals will be strongly connected to the scale of use and acceptance of fuel cells in vehicles. In addition, the efficiency of hydrogen production and delivery is significantly lower than the well-to-tank efficiency of many conventional fuels. It was concluded that an adequate hydrogen infrastructure will be required for the mining industry

  1. Two dimensional simulation of direct methanol fuel cell : a new (embedded) type of current collectors

    OpenAIRE

    Kulikovsky, A. A.; Divisek, J.; Kornyshev, Yu. M.

    2000-01-01

    A two-dimensional numerical model of the direct methanol fuel cell with gas fuel is developed. Simulation of the cell with current collectors of conventional geometry reveal the formation of fuel-depleted, "shaded" regions in the cathode and anode catalyst layers. These regions are positioned in front of current collectors, farther from the gas channel windows. Another disadvantage of the conventional geometry is the concentration of electron current at the edges of current collectors. Based ...

  2. Two-phase flow modeling for the cathode side of a Polymer electrolyte fuel cell

    NARCIS (Netherlands)

    Qin, C.; Rensink, D.; Fell, S.; Hassanizadeh, S.M.

    2012-01-01

    Liquid water flooding in micro gas channels is an important issue in the water management of polymer electrolyte fuel cells (PEFCs). However, in most previous numerical studies liquid water transport in the gas channels (GC) has been simplified by the mist flow assumption. In this work, we present a

  3. Effect of channel arrangement on fluid flow in PEMFC flow field using serpentine channel system with trapezoidal cross-section

    Energy Technology Data Exchange (ETDEWEB)

    Sun, L.; Oosthuizen, P.H. [Queen' s Univ., Kingston, ON (Canada). Dept. of Mechanical and Materials Engineering; McAuley, K.B. [Queen' s Univ., Kingston, ON (Canada). Dept. of Chemical Engineering

    2005-07-01

    Developments in Computational Flow Dynamics (CFD) software have meant that Proton Exchange Membrane Fuel Cell (PEMFC) modelling is now able to include cell components such as gas channels and porous diffusion layers. This paper discussed a numerical model which was developed to study air flow in the flow plate and gas diffusion layer assembly on the cathode side of a PEMFC. The flow plate in this fuel cell often has serpentine channels, and the porous layer is adjacent to the flow plate in order to diffuse the air to the catalyst layer. Flow crossover of air through the porous diffusion layer from one part of the channel to another can occur as a result of pressure differences between different parts of the channel. The numerical study was undertaken to compare the cases of a single channel and 2 parallel channels, with the channels having a trapezoidal cross-sectional shape. The objective of the study was to examine the effect of the flow plate geometry on the basic fluid flow through the plate. Flow was assumed to be 3-dimensional, steady, incompressible, isothermal and single-phase. The flow through the porous diffusion layer was described using the Darcy model. Dimensionless governing equations were solved using FIDAP, a commercial CFD solver. The results indicate that single channel systems have a greater maximum flow rate difference than the parallel channel systems under the conditions considered in the experiment. In addition, the size ratio R of trapezoidal cross-sectional shape has a significant effect on the flow crossover and pressure variation in the flow field. 16 refs., 15 figs.

  4. Oxy-fuel combustion of solid fuels

    DEFF Research Database (Denmark)

    Toftegaard, Maja Bøg; Brix, Jacob; Jensen, Peter Arendt

    2010-01-01

    Oxy-fuel combustion is suggested as one of the possible, promising technologies for capturing CO2 from power plants. The concept of oxy-fuel combustion is removal of nitrogen from the oxidizer to carry out the combustion process in oxygen and, in most concepts, recycled flue gas to lower the flame...... temperature. The flue gas produced thus consists primarily of carbon dioxide and water. Much research on the different aspects of an oxy-fuel power plant has been performed during the last decade. Focus has mainly been on retrofits of existing pulverized-coal-fired power plant units. Green-field plants which...... provide additional options for improvement of process economics are however likewise investigated. Of particular interest is the change of the combustion process induced by the exchange of carbon dioxide and water vapor for nitrogen as diluent. This paper reviews the published knowledge on the oxy-fuel...

  5. The first self-sustainable microbial fuel cell stack.

    Science.gov (United States)

    Ledezma, Pablo; Stinchcombe, Andrew; Greenman, John; Ieropoulos, Ioannis

    2013-02-21

    This study reports for the first time on the development of a self-sustainable microbial fuel cell stack capable of self-maintenance (feeding, hydration, sensing & reporting). Furthermore, the stack system is producing excess energy, which can be used for improved functionality. The self-maintenance is performed by the stack powering single and multi-channel peristaltic pumps.

  6. Experimental Study and Comparison of Various Designs of Gas Flow Fields to PEM Fuel Cells and Cell Stack Performance

    OpenAIRE

    Liu, Hong; Li, Peiwen; Juarez-Robles, Daniel; Wang, Kai; Hernandez-Guerrero, Abel

    2014-01-01

    In this study, a significant number of experimental tests to proton exchange membrane (PEM) fuel cells were conducted to investigate the effect of gas flow fields on fuel cell performance. Graphite plates with various flow field or flow channel designs, from literature survey and also novel designs by the authors, were used for the PEM fuel cell assembly. The fabricated fuel cells have an effective membrane area of 23.5 cm2. The results showed that the serpentine flow channel design is still ...

  7. Experimental study and comparison of various designs of gas flow fields to PEM fuel cells and cell stack performance

    OpenAIRE

    Hong eLiu; Peiwen eLi; Daniel eJuarez-Robles; Kai eWang; Abel eHernandez-Guerrero

    2014-01-01

    In this study, a significant number of experimental tests to PEM fuel cells were conducted to investigate the effect of gas flow fields on fuel cell performance. Graphite plates with various flow field or flow channel designs, from literature survey and also novel designs by the authors, were used for the PEM fuel cell assembly. The fabricated fuel cells all have an effective membrane area of 23.5 cm2. The results showed that the serpentine flow channel design is still favorable, giving the b...

  8. Cumulative Damage Fraction Evaluation for the Sub-channel Blockage Accident in PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Jin; Chang, Won-Pyo; Ha, Ki-Suk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    In determining a safety concern of sub-assembly, the CDF or life fraction, is very useful for predicting pins failure within sub-assembly that are subjected to creep damage at elevated temperatures and has been accepted as a means for predicting fuel pin failure in SFR. In particular, the sub-channels inside a fuel assembly in Sodium cooled Fast Reactor (SFR) may partially be blocked by an ingression of damaged fuel debris or foreign obstacles into fuel assembly due to the geometrically compact design of the core fuel pin arrangement. When the partial blockage occurs, sodium coolant flow would be disturbed in the vicinity of the blockage, and the affected flow could lead to a high local coolant temperature. The cladding breaching is assumed to occur when the CDF exceeds 1.0 and it is required that the CDF be below 1.0 to avoid the creep rupture of the cladding tube in fuel pin design. It is, therefore, important to evaluate the CDF of the fuel pins in an assembly of the 150MWe Prototype SFR. The objective of this paper is to predict the CDF of fuel pin within the hottest assembly which is designed in KAERI when the sub-channel blockage accident occurs. For the preliminary analysis, the CDF was calculated in the nominal condition without the hot channel factor. The evaluation of cumulative damage fraction was carried out for 6 sub-channel blockages in 150MWe Prototype Sodium cooled Fast Reactor (PGSFR) using the MATRA-LMR-FB code. The fuel peak temperature of the hottest pin was about 605 .deg. C and the CDF value obtained from the hottest pin during 1160 effective full power days (EFPDs) is 0.006, which means that fuel pins have large safety margins against breaching.

  9. Channel Access in Erlang

    Energy Technology Data Exchange (ETDEWEB)

    Nicklaus, Dennis J. [Fermilab

    2013-10-13

    We have developed an Erlang language implementation of the Channel Access protocol. Included are low-level functions for encoding and decoding Channel Access protocol network packets as well as higher level functions for monitoring or setting EPICS process variables. This provides access to EPICS process variables for the Fermilab Acnet control system via our Erlang-based front-end architecture without having to interface to C/C++ programs and libraries. Erlang is a functional programming language originally developed for real-time telecommunications applications. Its network programming features and list management functions make it particularly well-suited for the task of managing multiple Channel Access circuits and PV monitors.

  10. Fuel safety research 1999

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-07-01

    In April 1999, the Fuel Safety Research Laboratory was newly established as a result of reorganization of the Nuclear Safety Research Center, JAERI. The laboratory was organized by combining three laboratories, the Reactivity Accident Laboratory, the Fuel Reliability Laboratory, and a part of the Sever Accident Research Laboratory. Consequently, the Fuel Safety Research Laboratory is now in charge of all the fuel safety research in JAERI. Various types of experimental and analytical researches are conducted in the laboratory by using the unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and hot cells in JAERI. The laboratory consists of five research groups corresponding to each research fields. They are; (a) Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). (b) Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). (c) Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). (d) Research group of fuel behavior analysis (FEMAXI group). (e) Research group of FP release/transport behavior from irradiated fuel (VEGA group). This report summarizes the outline of research activities and major outcomes of the research executed in 1999 in the Fuel Safety Research Laboratory. (author)

  11. Fuel related risks; Braenslerisker

    Energy Technology Data Exchange (ETDEWEB)

    Englund, Jessica; Sernhed, Kerstin; Nystroem, Olle; Graveus, Frank (Grontmij AB, (Sweden))

    2012-02-15

    The project, within which this work report was prepared, aimed to complement the Vaermeforsk publication 'Handbook of fuels' on fuel related risks and measures to reduce the risks. The fuels examined in this project where the fuels included in the first version of the handbook from 2005 plus four additional fuels that will be included in the second and next edition of the handbook. Following fuels were included: woodfuels (sawdust, wood chips, powder, briquettes), slash, recycled wood, salix, bark, hardwood, stumps, straw, reed canary grass, hemp, cereal, cereal waste, olive waste, cocoa beans, citrus waste, shea, sludge, forest industrial sludge, manure, Paper Wood Plastic, tyre, leather waste, cardboard rejects, meat and bone meal, liquid animal and vegetable wastes, tall oil pitch, peat, residues from food industry, biomal (including slaughterhouse waste) and lignin. The report includes two main chapters; a general risk chapter and a chapter of fuel specific risks. The first one deals with the general concept of risk, it highlights laws and rules relevant for risk management and it discuss general risks that are related to the different steps of fuel handling, i.e. unloading, storing, processing the fuel, transportation within the facility, combustion and handling of ashes. The information that was used to produce this chapter was gathered through a literature review, site visits, and the project group's experience from risk management. The other main chapter deals with fuel-specific risks and the measures to reduce the risks for the steps of unloading, storing, processing the fuel, internal transportation, combustion and handling of the ashes. Risks and measures were considered for all the biofuels included in the second version in the handbook of fuels. Information about the risks and risk management was gathered through interviews with people working with different kinds of fuels in electricity and heat plants in Sweden. The information from

  12. Thermal assessment of Shippingport pressurized water reactor blanket fuel assemblies within a multi-canister overpack within the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-09

    A series of analyses were performed to assess the thermal performance characteristics of the Shippingport Pressurized Water Reactor Core 2 Blanket Fuel Assemblies as loaded within a Multi-Canister Overpack within the Canister Storage Building. A two-dimensional finite element was developed, with enough detail to model the individual fuel plates: including the fuel wafers, cladding, and flow channels.

  13. An Insight to Covert Channels

    OpenAIRE

    Salwan, Nitish; Singh, Sandeep; Arora, Suket; Singh, Amarpreet

    2013-01-01

    This paper presents an overview of different concepts regarding covert channels. It discusses the various classifications and the detailing of various fields used to manipulate for the covert channel execution.Different evaluation criterias are presented for measuring the strength of covert channels. The defenses and prevention schemes for this covert channel will also be discussed. This paper also discuss about an advanced timing channel i.e.Temperature Based Covert Channel.

  14. Channel Choice: A Literature Review

    DEFF Research Database (Denmark)

    Østergaard Madsen, Christian; Kræmmergaard, Pernille

    2015-01-01

    The channel choice branch of e-government studies citizens’ and businesses’ choice of channels for interacting with government, and how government organizations can integrate channels and migrate users towards the most cost-efficient channels. In spite of the valuable contributions offered no sys...... no systematic overview exist of channel choice. We present a literature review of channel choice studies in government to citizen context identifying authors, countries, methods, concepts, units of analysis, and theories, and offer suggestionsfor future studies....

  15. Direct Fuel Injector Temporal Measurements

    Science.gov (United States)

    2014-10-01

    optimize engine performance and emissions. Fuel injectors contain an actuator, pintle (or needle), and nozzle. The most common actuator is a solenoid ...Introduction Fuel injectors have a long history in metering fuel in modern engines by either port fuel injection (PFI) or direct fuel injection (DFI...Compared with a carburetor, fuel injectors have more accurate fuel delivering capability, thus giving engineers and technicians more flexibility to

  16. Cooperative gating between ion channels.

    Science.gov (United States)

    Choi, Kee-Hyun

    2014-01-01

    Cooperative gating between ion channels, i.e. the gating of one channel directly coupled to the gating of neighboring channels, has been observed in diverse channel types at the single-channel level. Positively coupled gating could enhance channel-mediated signaling while negative coupling may effectively reduce channel gating noise. Indeed, the physiological significance of cooperative channel gating in signal transduction has been recognized in several in vivo studies. Moreover, coupled gating of ion channels was reported to be associated with some human disease states. In this review, physiological roles for channel cooperativity and channel clustering observed in vitro and in vivo are introduced, and stimulation-induced channel clustering and direct channel cross linking are suggested as the physical mechanisms of channel assembly. Along with physical clustering, several molecular mechanisms proposed as the molecular basis for functional coupling of neighboring channels are covered: permeant ions as a channel coupling mediator, concerted channel activation through the membrane, and allosteric mechanisms. Also, single-channel analysis methods for cooperative gating such as the binomial analysis, the variance analysis, the conditional dwell time density analysis, and the maximum likelihood fitting analysis are reviewed and discussed.

  17. Sensing with Ion Channels

    CERN Document Server

    Martinac, Boris

    2008-01-01

    All living cells are able to detect and translate environmental stimuli into biologically meaningful signals. Sensations of touch, hearing, sight, taste, smell or pain are essential to the survival of all living organisms. The importance of sensory input for the existence of life thus justifies the effort made to understand its molecular origins. Sensing with Ion Channels focuses on ion channels as key molecules enabling biological systems to sense and process the physical and chemical stimuli that act upon cells in their living environment. Its aim is to serve as a reference to ion channel specialists and as a source of new information to non specialists who want to learn about the structural and functional diversity of ion channels and their role in sensory physiology.

  18. Covert Channels within IRC

    Science.gov (United States)

    2011-03-24

    Communications ....................................... 2 1.3 Steganography and Covert Channels .......................................................... 3...Internet Relay Chat ..................................................................................... 7 2.2 Steganography ...13 2.2.2 Encrypted Steganographic Systems .............................................. 15 2.2.3 Text-Based Steganography

  19. Channelized Streams in Iowa

    Data.gov (United States)

    Iowa State University GIS Support and Research Facility — This draft dataset consists of all ditches or channelized pieces of stream that could be identified using three input datasets; namely the1:24,000 National...

  20. Authentication over Noisy Channels

    CERN Document Server

    Lai, Lifeng; Poor, H Vincent

    2008-01-01

    In this work, message authentication over noisy channels is studied. The model developed in this paper is the authentication theory counterpart of Wyner's wiretap channel model. Two types of opponent attacks, namely impersonation attacks and substitution attacks, are investigated for both single message and multiple message authentication scenarios. For each scenario, information theoretic lower and upper bounds on the opponent's success probability are derived. Remarkably, in both scenarios, lower and upper bounds are shown to match, and hence the fundamental limit of message authentication over noisy channels is fully characterized. The opponent's success probability is further shown to be smaller than that derived in the classic authentication model in which the channel is assumed to be noiseless. These results rely on a proposed novel authentication scheme in which key information is used to provide simultaneous protection again both types of attacks.

  1. Cl- channels in apoptosis

    DEFF Research Database (Denmark)

    Wanitchakool, Podchanart; Ousingsawat, Jiraporn; Sirianant, Lalida

    2016-01-01

    , and cystic fibrosis transmembrane conductance regulator (CFTR) in cellular apoptosis. LRRC8A-E has been identified as a volume-regulated anion channel expressed in many cell types. It was shown to be required for regulatory and apoptotic volume decrease (RVD, AVD) in cultured cell lines. Its presence also......(-) channels or as regulators of other apoptotic Cl(-) channels, such as LRRC8. CFTR has been known for its proapoptotic effects for some time, and this effect may be based on glutathione release from the cell and increase in cytosolic reactive oxygen species (ROS). Although we find that CFTR is activated...... by cell swelling, it is possible that CFTR serves RVD/AVD through accumulation of ROS and activation of independent membrane channels such as ANO6. Thus activation of ANO6 will support cell shrinkage and induce additional apoptotic events, such as membrane phospholipid scrambling....

  2. Volume Regulated Channels

    DEFF Research Database (Denmark)

    Klausen, Thomas Kjær

    - serves a multitude of functions in the mammalian cell, regulating the membrane potential (Em), cell volume, protein activity and the driving force for facilitated transporters giving Cl- and Cl- channels a major potential of regulating cellular function. These functions include control of the cell cycle...... of volume perturbations evolution have developed system of channels and transporters to tightly control volume homeostasis. In the past decades evidence has been mounting, that the importance of these volume regulated channels and transporters are not restricted to the defense of cellular volume......, controlled cell death and cellular migration. Volume regulatory mechanisms has long been in focus for regulating cellular proliferation and my thesis work have been focusing on the role of Cl- channels in proliferation with specific emphasis on ICl, swell. Pharmacological blockage of the ubiquitously...

  3. Fuel cell generator with fuel electrodes that control on-cell fuel reformation

    Science.gov (United States)

    Ruka, Roswell J.; Basel, Richard A.; Zhang, Gong

    2011-10-25

    A fuel cell for a fuel cell generator including a housing including a gas flow path for receiving a fuel from a fuel source and directing the fuel across the fuel cell. The fuel cell includes an elongate member including opposing first and second ends and defining an interior cathode portion and an exterior anode portion. The interior cathode portion includes an electrode in contact with an oxidant flow path. The exterior anode portion includes an electrode in contact with the fuel in the gas flow path. The anode portion includes a catalyst material for effecting fuel reformation along the fuel cell between the opposing ends. A fuel reformation control layer is applied over the catalyst material for reducing a rate of fuel reformation on the fuel cell. The control layer effects a variable reformation rate along the length of the fuel cell.

  4. Chloride channels in stroke

    Institute of Scientific and Technical Information of China (English)

    Ya-ping ZHANG; Hao ZHANG; Dayue Darrel DUAN

    2013-01-01

    Vascular remodeling of cerebral arterioles,including proliferation,migration,and apoptosis of vascular smooth muscle cells (VSMCs),is the major cause of changes in the cross-sectional area and diameter of the arteries and sudden interruption of blood flow or hemorrhage in the brain,ie,stroke.Accumulating evidence strongly supports an important role for chloride (Clˉ) channels in vascular remodeling and stroke.At least three Clˉ channel genes are expressed in VSMCs:1) the TMEM16A (or Ano1),which may encode the calcium-activated Clˉ channels (CACCs); 2) the CLC-3 Clˉ channel and Clˉ/H+ antiporter,which is closely related to the volume-regulated Clˉ channels (VRCCs); and 3) the cystic fibrosis transmembrane conductance regulator (CFTR),which encodes the PKA-and PKC-activated Clˉ channels.Activation of the CACCs by agonist-induced increase in intracellular Ca2+ causes membrane depolarization,vasoconstriction,and inhibition of VSMC proliferation.Activation of VRCCs by cell volume increase or membrane stretch promotes the production of reactive oxygen species,induces proliferation and inhibits apoptosis of VSMCs.Activation of CFTR inhibits oxidative stress and may prevent the development of hypertension.In addition,Clˉ current mediated by gammaaminobutyric acid (GABA) receptor has also been implicated a role in ischemic neuron death.This review focuses on the functional roles of Clˉ channels in the development of stroke and provides a perspective on the future directions for research and the potential to develop Clˉ channels as new targets for the prevention and treatment of stroke.

  5. Solid electrolytic fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Sakai, Masayasu; Yamauchi, Yasuhiro; Kamisaka, Mitsuo; Notomi, Kei.

    1989-04-21

    Concerning a solid electrolytic fuel cell with a gas permeable substrate pipe, a fuel electrode installed on this substrate pipe and an air electrode which is laminated on this fuel electrode with the electrolyte in between, the existing fuel cell of this kind uses crystals of CaMnO3, etc. for the material of the air electrode, but its electric resistance is big and in order to avert this, it is necessary to make the film thickness of the air electrode big. However, in such a case, the entry of the air into its inside worsens and the cell performance cannot develop satisfactorily. In view of the above, in order to obtain a high performance solid electrolytic fuel cell which can improve electric conductivity without damaging diffusion rate of the air, this invention proposes with regard to the aforementioned solid electrolytic fuel cell to install a heat resistant and conductive member inside the above air electrode. 6 figs.

  6. HTPEM Fuel Cell Impedance

    DEFF Research Database (Denmark)

    Vang, Jakob Rabjerg

    As part of the process to create a fossil free Denmark by 2050, there is a need for the development of new energy technologies with higher efficiencies than the current technologies. Fuel cells, that can generate electricity at higher efficiencies than conventional combustion engines, can...... potentially play an important role in the energy system of the future. One of the fuel cell technologies, that receives much attention from the Danish scientific community is high temperature proton exchange membrane (HTPEM) fuel cells based on polybenzimidazole (PBI) with phosphoric acid as proton conductor....... This type of fuel cell operates at higher temperature than comparable fuel cell types and they distinguish themselves by high CO tolerance. Platinum based catalysts have their efficiency reduced by CO and the effect is more pronounced at low temperature. This Ph.D. Thesis investigates this type of fuel...

  7. Toward sustainable fuel cells

    DEFF Research Database (Denmark)

    Stephens, Ifan; Rossmeisl, Jan; Chorkendorff, Ib

    2016-01-01

    A quarter of humanity's current energy consumption is used for transportation (1). Low-temperature hydrogen fuel cells offer much promise for replacing this colossal use of fossil fuels with renewables; these fuel cells produce negligible emissions and have a mileage and filling time equal...... to a regular gasoline car. However, current fuel cells require 0.25 g of platinum (Pt) per kilowatt of power (2) as catalysts to drive the electrode reactions. If the entire global annual production of Pt were devoted to fuel cell vehicles, fewer than 10 million vehicles could be produced each year, a mere 10......% of the annual automotive vehicle production. Lowering the Pt loading in a fuel cell to a sustainable level requires the reactivity of Pt to be tuned so that it accelerates oxygen reduction more effectively (3). Two reports in this issue address this challenge (4, 5)....

  8. Spiral cooled fuel nozzle

    Energy Technology Data Exchange (ETDEWEB)

    Fox, Timothy; Schilp, Reinhard

    2012-09-25

    A fuel nozzle for delivery of fuel to a gas turbine engine. The fuel nozzle includes an outer nozzle wall and a center body located centrally within the nozzle wall. A gap is defined between an inner wall surface of the nozzle wall and an outer body surface of the center body for providing fuel flow in a longitudinal direction from an inlet end to an outlet end of the fuel nozzle. A turbulating feature is defined on at least one of the central body and the inner wall for causing at least a portion of the fuel flow in the gap to flow transverse to the longitudinal direction. The gap is effective to provide a substantially uniform temperature distribution along the nozzle wall in the circumferential direction.

  9. FACTAR 2.0 code validation

    Energy Technology Data Exchange (ETDEWEB)

    Middleton, P.B.; Rock, R.C.K.; Wadsworth, S.L. [Ontario Hydro, Reactor Safety and Operational Analysis Dept., Toronto, Ontario (Canada)

    1997-07-01

    The FACTAR code models the thermal and mechanical behaviour of a CANDU fuel channel under degraded cooling conditions. FACTAR is currently undergoing a process of validation against various data sets in order to qualify its use in nuclear safety analysis. This paper outlines the methodology being followed in this effort. The BTF-104 and BTF-105A tests, conducted at Chalk River Laboratories, have been chosen as the first in reactor tests to be used for FACTAR validation. The BTF experiments were designed to represent CANDU fuel behaviour under typical large LOCA conditions. The two tests are summarized briefly, and the results of code comparisons to experimental data are outlined. The comparisons demonstrate that FACTAR is able to accurately predict the values of selected key parameters. As anticipated in the validation plan, further work is required to fully quantify simulation biases for all parameters of interest. (author)

  10. Countercurrent flow limited (CCFL) heat flux in the high flux isotope reactor (HFIR) fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Ruggles, A.E.

    1990-10-12

    The countercurrent flow (CCF) performance in the fuel element region of the HFIR is examined experimentally and theoretically. The fuel element consists of two concentric annuli filled with aluminum clad fuel plates of 1.27 mm thickness separated by 1.27 mm flow channels. The plates are curved as they go radially outward to accomplish constant flow channel width and constant metal-to-coolant ratio. A full-scale HFIR fuel element mock-up is studied in an adiabatic air-water CCF experiment. A review of CCF models for narrow channels is presented along with the treatment of CCFs in system of parallel channels. The experimental results are related to the existing models and a mechanistic model for the annular'' CCF in a narrow channel is developed that captures the data trends well. The results of the experiment are used to calculate the CCFL heat flux of the HFIR fuel assembly. It was determined that the HFIR fuel assembly can reject 0.62 Mw of thermal power in the CCFL situation. 31 refs., 17 figs.

  11. Fracture channel waves

    Science.gov (United States)

    Nihei, Kurt T.; Yi, Weidong; Myer, Larry R.; Cook, Neville G. W.; Schoenberg, Michael

    1999-03-01

    The properties of guided waves which propagate between two parallel fractures are examined. Plane wave analysis is used to obtain a dispersion equation for the velocities of fracture channel waves. Analysis of this equation demonstrates that parallel fractures form an elastic waveguide that supports two symmetric and two antisymmetric dispersive Rayleigh channel waves, each with particle motions and velocities that are sensitive to the normal and tangential stiffnesses of the fractures. These fracture channel waves degenerate to shear waves when the fracture stiffnesses are large, to Rayleigh waves and Rayleigh-Lamb plate waves when the fracture stiffnesses are low, and to fracture interface waves when the fractures are either very closely spaced or widely separated. For intermediate fracture stiffnesses typical of fractured rock masses, fracture channel waves are dispersive and exhibit moderate to strong localization of guided wave energy between the fractures. The existence of these waves is examined using laboratory acoustic measurements on a fractured marble plate. This experiment confirms the distinct particle motion of the fundamental antisymmetric fracture channel wave (A0 mode) and demonstrates the ease with which a fracture channel wave can be generated and detected.

  12. Course on Ionic Channels

    CERN Document Server

    1986-01-01

    This book is based on a series of lectures for a course on ionic channels held in Santiago, Chile, on November 17-20, 1984. It is intended as a tutorial guide on the properties, function, modulation, and reconstitution of ionic channels, and it should be accessible to graduate students taking their first steps in this field. In the presentation there has been a deliberate emphasis on the spe­ cific methodologies used toward the understanding of the workings and function of channels. Thus, in the first section, we learn to "read" single­ channel records: how to interpret them in the theoretical frame of kinetic models, which information can be extracted from gating currents in re­ lation to the closing and opening processes, and how ion transport through an open channel can be explained in terms of fluctuating energy barriers. The importance of assessing unequivocally the origin and purity of mem­ brane preparations and the use of membrane vesicles and optical tech­ niques in the stUGY of ionic channels a...

  13. Liquid fuel cells.

    Science.gov (United States)

    Soloveichik, Grigorii L

    2014-01-01

    The advantages of liquid fuel cells (LFCs) over conventional hydrogen-oxygen fuel cells include a higher theoretical energy density and efficiency, a more convenient handling of the streams, and enhanced safety. This review focuses on the use of different types of organic fuels as an anode material for LFCs. An overview of the current state of the art and recent trends in the development of LFC and the challenges of their practical implementation are presented.

  14. Fuel safety research 2001

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-11-01

    The Fuel Safety Research Laboratory is in charge of research activity which covers almost research items related to fuel safety of water reactor in JAERI. Various types of experimental and analytical researches are being conducted by using some unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and the Reactor Fuel Examination Facility (RFEF) of JAERI. The research to confirm the safety of high burn-up fuel and MOX fuel under accident conditions is the most important item among them. The laboratory consists of following five research groups corresponding to each research fields; Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). Research group of fuel behavior under the normal operation cond