WorldWideScience

Sample records for cooled htpem fuel

  1. 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...

  2. Modeling and Implementation of a 1 kW, Air Cooled HTPEM Fuel Cell in a Hybrid Electrical Vehicle

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Ashworth, Leanne; Remón, Ian Natanael

    2008-01-01

    This work is a preliminary study of using the PBI-based, HTPEM fuel cell technology in automotive applications. This issue was investigated through computational modeling and an experimental investigation. A hybrid fuel cell system, consisting of a 1 kW stack and lead acid batteries, was implemen......This work is a preliminary study of using the PBI-based, HTPEM fuel cell technology in automotive applications. This issue was investigated through computational modeling and an experimental investigation. A hybrid fuel cell system, consisting of a 1 kW stack and lead acid batteries......, was implemented in a small electrical vehicle. A dynamic model was developed using Matlab-Simulink to describe the system characteristics, select operating conditions and to size system components. Preheating of the fuel cell stack with electrical resistors was investigated and found to be an unrealistic approach...

  3. Experimental study and modelling of degradation phenomena in HTPEM fuel cell stacks for use in CHP systems

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl

    2009-01-01

    Degradation phenomena in HTPEM fuel cells for use in CHP systems were investigated experimentally and by modelling. It was found that the two main degradation mechanisms in HTPEM fuel cells are carbon corrosion and Pt agglomeration. On basis of this conclusion a mechanistic model, describing...

  4. HT-PEM Fuel Cell System with Integrated Thermoelectric Exhaust Heat Recovery

    DEFF Research Database (Denmark)

    Gao, Xin

    This thesis presents two case studies on improving the efficiency and the loadfollowing capability of a high temperature polymer electrolyte membrane (HTPEM) fuel cell system by the application of thermoelectric (TE) devices. TE generators (TEGs) are harnessed to recover the system exhaust gas...... developed three-dimensional numerical model in ANSYS Fluent®. This thesis introduces the progress of this project in a cognitive order. The first chapter initially prepares the theory and characteristics of the fuel cell system and TE devices. Project motivations are conceived. Then similar studies existing...... power output on the subsystem design and performance were also systematically analyzed. The TEG subsystem configuration is optimized. The usefulness and convenience of the model are proved. TE coolers (TECs) are integrated into the methanol evaporator of the HT-PEM system for improving the whole system...

  5. Experimental Evaluation of a Pt-based Heat Exchanger Methanol Reformer for a HTPEM Fuel Cell Stack

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Kær, Søren Knudsen; Nielsen, Mads Pagh

    2008-01-01

    and automotive applications. Using a liquid hydrocarbon as e.g. methanol as the hydrogen carrier and reforming it to a hydrogen rich gas can solve some of these storage issues. The work presented here examines the use of a heat exchanger methanol reformer for use with a HTPEM fuel cell stack. Initial......Fuel cell systems running on pure hydrogen can efficiently produce electricity and heat for various applications, stationary and mobile. Storage volume can be problematic for stationary fuel cell systems with high run-time demands, but it is especially a challenge when dealing with mobile...

  6. Directly connected series coupled HTPEM fuel cell stacks to a Li-ion battery DC bus for a fuel cell electrical vehicle

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Ashworth, Leanne; Remón, Ian Natanael

    2008-01-01

    The work presented in this paper examines the use of pure hydrogen fuelled high temperature polymer electrolyte membrane (HTPEM) fuel cell stacks in an electrical car, charging a Li-ion battery pack. The car is equipped with two branches of two series coupled 1 kW fuel cell stacks which...... are connected directly parallel to the battery pack during operation. This enables efficient charging of the batteries for increased driving range. With no power electronics used, the fuel cell stacks follow the battery pack voltage, and charge the batteries passively. This saves the electrical and economical...... losses related to these components and their added system complexity. The new car battery pack consists of 23 Li-ion battery cells and the charging and discharging are monitored by a battery management system (BMS) which ensures safe operating conditions for the batteries. The direct connection...

  7. Diagnosis of CO Pollution in HTPEM Fuel Cell using Statistical Change Detection

    DEFF Research Database (Denmark)

    Jeppesen, Christian; Blanke, Mogens; Zhou, Fan

    2015-01-01

    The fuel cell technologies are advancing and maturing for commercial markets. However proper diagnostic tools needs to be developed in order to insure reliability and durability of fuel cell systems. This paper presents a design of a data driven method to detect CO content in the anode gas...... of a high temperature fuel cell. In this work the fuel cell characterization is based on an experimental equivalent electrical circuit, where model parameters are mapped as a function of the load current. The designed general likelihood ratio test detection scheme detects whether a equivalent electrical...... circuit parameter differ from the non-faulty operation. It is proven that the general likelihood ratio test detection scheme, with a very low probability of false alarm, can detect CO content in the anode gas of the fuel cell....

  8. Modeling of a HTPEM fuel cell using Adaptive Neuro-Fuzzy Inference Systems

    DEFF Research Database (Denmark)

    Justesen, Kristian Kjær; Andreasen, Søren Juhl; Sahlin, Simon Lennart

    2015-01-01

    In this work an Adaptive Neuro-Fuzzy Inference System (ANFIS) model of the voltage of a fuel cell is developed. The inputs of this model are the fuel cell temperature, current density and the carbon monoxide concentration of the anode supply gas. First an identification experiment which spans...... the expected operating range of the fuel cell is performed in a test station. The data from this experiment is then used to train ANFIS models with 2, 3, 4 and 5 membership functions. The performance of these models is then compared and it is found that using 3 membership functions provides the best compromise...

  9. 400 W High Temperature PEM Fuel Cell Stack Test

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Kær, Søren Knudsen

    2006-01-01

    This work demonstrates the operation of a 30 cell high temperature PEM (HTPEM) fuel cell stack. This prototype stack has been developed at the Institute of Energy Technology, Aalborg University, as a proof-of-concept for a low pressure cathode air cooled HTPEM stack. The membranes used are Celtec...

  10. Analysis of accelerated degradation of a HT-PEM fuel cell caused by cell reversal in fuel starvation condition

    DEFF Research Database (Denmark)

    Zhou, Fan; Andreasen, Søren Juhl; Kær, Søren Knudsen

    2015-01-01

    This paper reports an accelerated degradation test of a high temperature PEM fuel cell under repeated H2 starvation condition. The H2 stoichiometry is cycled between 3.0 and 0.8 every 2 min during the test. The experimental results show that the polarity of the fuel cell is reversed under H2......, there is only a slight decrease in open circuit voltage of the fuel cell which implies the membrane is not affected by the test. The electrochemical impedance spectrum measurement shows that the H2 starvation can cause significant increase in the ohmic resistance and charge transfer resistance. By looking...... starvation condition, and the cell performance indicated by cell voltage at H2 stoichiometry of 3.0 declines from 0.59 V to 0.41 V in 19 cycles. Since CO2 is detected in anode exhaust under H2 starvation condition, carbon corrosion is believed to be the reason for the degradation in this test. After the test...

  11. Preliminary design of a small-scale system for the conversion of biogas to electricity by HT-PEM fuel cell

    International Nuclear Information System (INIS)

    Birth, Torsten; Heineken, Wolfram; He, Ling

    2014-01-01

    In this work a novel concept for the decentralized conversion of biogas to electricity is introduced. It consists of five segments: gas supply, gas treatment, gas reforming, gas usage and post-combustion. The system was designed in a regional project called GREEN-FC. The project is dealing with a design study for the conversion of 1 m 3  h −1 biogas to electricity, based on equilibrium calculations for steam reforming and water–gas shift reaction in combination with CFD simulations. The simulation results revealed that the system converts methane fully and delivers a maximum yield of hydrogen with a low concentration of carbon monoxide, thus making it suitable for a high-temperature polymer–electrolyte membrane (HT-PEM) fuel cell. The calculated electrical efficiency of the novel process is approximately 40%. Another important result of this work is the modular prototype design, because the individual components of the prototype can be replaced. For example alternative reactors that convert biogas into hydrogen and other technologies that use hydrogen can be included. - Highlights: • We designed the GREEN-FC process for decentralized hydrogen production from biogas. • We determined optimal process conditions on chemical equilibrium calculations. • The design was evaluated by CFD simulations with chemical reactions included. • The electrical efficiency of the GREEN-FC process is approximately 40%. • The first industrial prototype should have investment costs of 5000 € kW −1

  12. EIS Characterization of the Poisoning Effects of CO and CO2 on a PBI based HT-PEM Fuel Cell

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Mosbæk, Rasmus; Vang, Jakob Rabjerg

    2010-01-01

    . All operational parameters as well as data acquisition are controlled by two LabView programs, running on two separate computers. The impedance spectrum of the fuel cell is recorded at different operating points and then an Equivalent Circuit (EC), proposed for modelling the cell impedance, is fitted...

  13. Vapor Delivery Systems for the Study of the Effects of Reformate Gas Impurities in HT-PEM Fuel Cells

    DEFF Research Database (Denmark)

    Araya, Samuel Simon; Kær, Søren Knudsen; Andreasen, Søren Juhl

    2011-01-01

    , impurities in the reformate gas produced from methanol steam reforming can affect the performance and durability of fuel cells. In this paper different vapor delivery systems, intended to assist in the study of the effects of some of the impurities, are described and compared with each other. A system based...... on a pump and electrically heated evaporator was found to be more suitable for the typical flow rates involved in the anode feed of an H3PO4/PBI based HT-PEMFC unit cell assembly. Test stations composed of vapor delivery systems and mass flow controllers for testing the effects of methanol slip, water vapor...

  14. 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.

  15. Cooling nuclear reactor fuel

    International Nuclear Information System (INIS)

    Porter, W.H.L.

    1975-01-01

    Reference is made to water or water/steam cooled reactors of the fuel cluster type. In such reactors it is usual to mount the clusters in parallel spaced relationship so that coolant can pass freely between them, the coolant being passed axially from one end of the cluster in an upward direction through the cluster and being effective for cooling under normal circumstances. It has been suggested, however, that in addition to the main coolant flow an auxiliary coolant flow be provided so as to pass laterally into the cluster or be sprayed over the top of the cluster. This auxiliary supply may be continuously in use, or may be held in reserve for use in emergencies. Arrangements for providing this auxiliary cooling are described in detail. (U.K.)

  16. Assessment of spent fuel cooling

    International Nuclear Information System (INIS)

    Ibarra, J.G.; Jones, W.R.; Lanik, G.F.

    1997-01-01

    The paper presents the methodology, the findings, and the conclusions of a study that was done by the Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data (AEOD) on loss of spent fuel pool cooling. The study involved an examination of spent fuel pool designs, operating experience, operating practices, and procedures. AEOD's work was augmented in the area of statistics and probabilistic risk assessment by experts from the Idaho Nuclear Engineering Laboratory. Operating experience was integrated into a probabilistic risk assessment to gain insight on the risks from spent fuel pools

  17. Dynamic Model of High Temperature PEM Fuel Cell Stack Temperature

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Kær, Søren Knudsen

    2007-01-01

    cathode air cooled 30 cell HTPEM fuel cell stack developed at the Institute of Energy Technology at Aalborg University. This fuel cell stack uses PEMEAS Celtec P-1000 membranes, runs on pure hydrogen in a dead end anode configuration with a purge valve. The cooling of the stack is managed by running......The present work involves the development of a model for predicting the dynamic temperature of a high temperature PEM (HTPEM) fuel cell stack. The model is developed to test different thermal control strategies before implementing them in the actual system. The test system consists of a prototype...... the stack at a high stoichiometric air flow. This is possible because of the PBI fuel cell membranes used, and the very low pressure drop in the stack. The model consists of a discrete thermal model dividing the stack into three parts: inlet, middle and end and predicting the temperatures in these three...

  18. Control and experimental characterization of a methanol reformer for a 350 W high temperature polymer electrolyte membrane fuel cell system

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Kær, Søren Knudsen; Sahlin, Simon Lennart

    2013-01-01

    is the water and methanol mixture fuel flow and the burner fuel/air ratio and combined flow. An experimental setup is presented capable of testing the methanol reformer used in the Serenergy H3 350 Mobile Battery Charger; a high temperature polymer electrolyte membrane (HTPEM) fuel cell system......This work presents a control strategy for controlling the methanol reformer temperature of a 350 W high temperature polymer electrolyte membrane fuel cell system, by using a cascade control structure for reliable system operation. The primary states affecting the methanol catalyst bed temperature....... The experimental system consists of a fuel evaporator utilizing the high temperature waste gas from the cathode air cooled 45 cell HTPEM fuel cell stack. The fuel cells used are BASF P1000 MEAs which use phosphoric acid doped polybenzimidazole membranes. The resulting reformate gas output of the reformer system...

  19. Fuel arrangement for high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Tobin, J.M.

    1978-01-01

    Disclosed is a fuel arrangement for a high temperature gas cooled reactor including fuel assemblies with separate directly cooled fissile and fertile fuel elements removably inserted in an elongated moderator block also having a passageway for control elements

  20. A combined capillary cooling system for cooling fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Ana Paula; Pelizza, Pablo Rodrigo; Galante, Renan Manozzo; Bazzo, Edson [Universidade Federal de Santa Catarina (LabCET/UFSC), Florianopolis, SC (Brazil). Dept. de Engenharia Mecanica. Lab. de Combustao e Engenharia de Sistemas Termicos], Emails: ana@labcet.ufsc.br, pablo@labcet.ufsc.br, renan@labcet.ufsc.br, ebazzo@emc.ufsc.br

    2010-07-01

    The operation temperature control has an important influence over the PEMFC (Proton Exchange Membrane Fuel Cell) performance. A two-phase heat transfer system is proposed as an alternative for cooling and thermal control of PEMFC. The proposed system consists of a CPL (Capillary Pumped Loop) connected to a set of constant conductance heat pipes. In this work ceramic wick and stainless mesh wicks have been used as capillary structure of the CPL and heat pipes, respectively. Acetone has been used as the working fluid for CPL and deionized water for the heat pipes. Experimental results of three 1/4 inch stainless steel outlet diameter heats pipes and one CPL have been carried out and presented in this paper. Further experiments are planned coupling the proposed cooling system to a module which simulates the fuel cell. (author)

  1. Performance Degradation Tests of Phosphoric Acid Doped Polybenzimidazole Membrane Based High Temperature Polymer Electrolyte Membrane Fuel Cells

    DEFF Research Database (Denmark)

    Zhou, Fan; Araya, Samuel Simon; Grigoras, Ionela

    2015-01-01

    Degradation tests of two phosphoric acid (PA) doped PBI membrane based HT-PEM fuel cells were reported in this paper to investigate the effects of start/stop and the presence of methanol in the fuel to the performance degradation of the HT-PEM fuel cell. Continuous tests with pure dry H2 and meth...

  2. Modelling of a High Temperature PEM Fuel Cell Stack using Electrochemical Impedance Spectroscopy

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Jespersen, Jesper Lebæk; Kær, Søren Knudsen

    2008-01-01

    This work presents the development of an equivalent circuit model of a 65 cell high temperature PEM (HTPEM) fuel cell stack using Electrochemical Impedance Spectroscopy (EIS). The HTPEM fuel cell membranes used are PBI-based and uses phosphoric acid as proton conductor. The operating temperature...

  3. Modelling and Evaluation of Heating Strategies for High Temperature Polymer Electrolyte Membrane Fuel Cell Stacks

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Kær, Søren Knudsen

    2008-01-01

    Experiments were conducted on two different cathode air cooled high temperature PEM (HTPEM) fuel cell stacks; a 30 cell 400W prototype stack using two bipolar plates per cell, and a 65 cell 1 kW commercial stack using one bipolar plate per cell. The work seeks to examine the use of different...... model to simulate the temperature development of a fuel cell stack during heating can be used for assistance in system and control design. The heating strategies analyzed and tested reduced the startup time of one of the fuel cell stacks from 1 h to about 6 min....

  4. Effects of cooling time on a closed LWR fuel cycle

    International Nuclear Information System (INIS)

    Arnold, R. P.; Forsberg, C. W.; Shwageraus, E.

    2012-01-01

    In this study, the effects of cooling time prior to reprocessing spent LWR fuel has on the reactor physics characteristics of a PWR fully loaded with homogeneously mixed U-Pu or U-TRU oxide (MOX) fuel is examined. A reactor physics analysis was completed using the CASM04e code. A void reactivity feedback coefficient analysis was also completed for an infinite lattice of fresh fuel assemblies. Some useful conclusions can be made regarding the effect that cooling time prior to reprocessing spent LWR fuel has on a closed homogeneous MOX fuel cycle. The computational analysis shows that it is more neutronically efficient to reprocess cooled spent fuel into homogeneous MOX fuel rods earlier rather than later as the fissile fuel content decreases with time. Also, the number of spent fuel rods needed to fabricate one MOX fuel rod increases as cooling time increases. In the case of TRU MOX fuel, with time, there is an economic tradeoff between fuel handling difficulty and higher throughput of fuel to be reprocessed. The void coefficient analysis shows that the void coefficient becomes progressively more restrictive on fuel Pu content with increasing spent fuel cooling time before reprocessing. (authors)

  5. Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    Topics covered during the 'Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment' were as follows: Selection of constructions and materials, fuel element development concepts; Fabrication of spherical coated fuel particles and fuel element on their base; investigation of fuel properties; Spent fuel treatment and storage; Head-end processing of HTGR fuel elements; investigation of HTGR fuel regeneration process; applicability of gas-fluorine technology of regeneration of spent HTGR fuel elements.

  6. Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment

    International Nuclear Information System (INIS)

    1985-01-01

    Topics covered during the 'Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment' were as follows: Selection of constructions and materials, fuel element development concepts; Fabrication of spherical coated fuel particles and fuel element on their base; investigation of fuel properties; Spent fuel treatment and storage; Head-end processing of HTGR fuel elements; investigation of HTGR fuel regeneration process; applicability of gas-fluorine technology of regeneration of spent HTGR fuel elements

  7. 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.

  8. Fuel assembly cooling experience at the FFTF/IEM cell

    International Nuclear Information System (INIS)

    McGuinness, P.W.

    1985-01-01

    In the Fast Flux Test Facility (FFTF), sodium wetted irradiated fuel assemblies are discharged to the Interim Examination and Maintenance (IEM) Cell for disassembly and post-irradiation examination in an inert argon atmosphere. While in the IEM Cell, fuel assemblies are cooled by the IEM Cell Subassembly Cooling System. This paper describes the cooling system design, performance, and lessons learned, including a discussion of two overtemperature incidents. 2 refs., 6 figs

  9. Experimental Analysis of the Effects of CO and CO2 on High Temperature PEM Fuel Cell Performance using Electrochemical Impedance Spectroscopy

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Vang, Jakob Rabjerg

    2010-01-01

    The use of high temperature PEM (HTPEM) fuel cells running on reformate gas shows comparable performance to HTPEM fuel cells running on pure hydrogen, even when running at high levels of CO, as long as high operating temperatures are ensured. The increased operating temperatures of these types of...

  10. Evaporative cooling in polymer electrolyte fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Shimotori, S; Sonai, A [Toshiba Corp. Tokyo (Japan)

    1996-06-05

    The concept of the evaporative cooling for the internally humidified PEFC was confirmed by the experiment. The evaporative cooling rates at the anode and the cathode were mastered under the various temperatures and air utilizations. At a high temperature the proportion of the evaporative cooling rate to the heat generation rate got higher, the possibility of the evaporative cooling was demonstrated. 2 refs., 7 figs., 1 tab.

  11. Design and Control of High Temperature PEM Fuel Cell System

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl

    E-cient fuel cell systems have started to appear in many dierent commercial applications and large scale production facilities are already operating to supply fuel cells to support an ever growing market. Fuel cells are typically considered to replace leadacid batteries in applications where...... to conventional PEM fuel cells, that use liquid water as a proton conductor and thus operate at temperatures below 100oC. The HTPEM fuel cell membrane in focus in this work is the BASF Celtec-P polybenzimidazole (PBI) membrane that uses phosphoric acid as a proton conductor. The absence of water in the fuel cells...... enables the use of designing cathode air cooled stacks greatly simplifying the fuel cell system and lowering the parasitic losses. Furthermore, the fuel impurity tolerance is signicantly improved because of the higher temperatures, and much higher concentrations of CO can be endured without performance...

  12. Post-accident cooling capacity analysis of the AP1000 passive spent fuel pool cooling system

    International Nuclear Information System (INIS)

    Su Xia

    2013-01-01

    The passive design is used in AP1000 spent fuel pool cooling system. The decay heat of the spent fuel is removed by heating-boiling method, and makeup water is provided passively and continuously to ensure the safety of the spent fuel. Based on the analysis of the post-accident cooling capacity of the spent fuel cooling system, it is found that post-accident first 72-hour cooling under normal refueling condition and emergency full-core offload condition can be maintained by passive makeup from safety water source; 56 hours have to be waited under full core refueling condition to ensure the safety of the core and the spent fuel pool. Long-term cooling could be conducted through reserved safety interface. Makeup measure is available after accident and limited operation is needed. Makeup under control could maintain the spent fuel at sub-critical condition. Compared with traditional spent fuel pool cooling system design, the AP1000 design respond more effectively to LOCA accidents. (authors)

  13. Development of a 400 W High Temperature PEM Fuel Cell Power Pack

    DEFF Research Database (Denmark)

    Schaltz, Erik; Jespersen, Jesper Lebæk; Rasmussen, Peter Omand

    2006-01-01

    reformer design because CO removal is not needed. A fuel like methanol would be a preferable choice for reforming when using HTPEM fuel cells because of its high energy density and low reforming temperatures. The thermal integration and use of HTPEM fuel cells with methanol reformers show promising results......When using pressurized hydrogen to fuel a fuel cell, much space is needed for fuel storage. This is undesirable especially with mobile or portable fuel cell systems, where refuelling also often is inconvenient. Using a reformed liquid carbonhydrate can reduce this fuel volume considerably. Nafion...... based low temperature PEM (LTPEM) fuel cells are very intolerant to reformate gas because of the presence of CO. PBI based high temperature PEM (HTPEM) fuel cells can operate stable at much higher CO concentrations. This makes the HTPEM very suitable for applications using a reformer, and could simplify...

  14. Cooling-time determination of spent fuel

    International Nuclear Information System (INIS)

    Hsue, S.T.; Hatcher, C.R.; Kaieda, K.

    1979-01-01

    Two methods to determine the cooling time using data from high-resolution gamma-ray measurements are discussed; one is useful when the irradiation history information is available, the other when the irradiation history is not available. We have applied both methods and found that the cooling time can be determined to within an average of 3% and 4.1%, respectively

  15. Development of design technology for dual-cooled fuel

    International Nuclear Information System (INIS)

    Kim, Hyung Kyu; Yoon, Kyung Ho; Lee, Young Ho

    2010-03-01

    Primary purpose of the project is to complete a basic design of the power uprating dual-cooled fuel's structural components for an actual use in the existing nuclear power plants. It also includes a basic design of the components of a dual-cooled fuel rod. To this end, during the three years of the first stage (2007.03.∼2010.02.), concepts and technical issues of the structural components such as a supporting structure, guide thimbles and instrumentation tube and the top and bottom end pieces were derived in order to comply with the functional requirements and design criteria of them. Basic design was carried out to resolve the issues by using analytical methods as well as experiments, and observed finally is that a structural compatibility of the designed dual-cooled fuel to the Korean Standard Nuclear Power Plant (OPR-1000). As for the dual-cooled fuel rod's components such as a plenum spring, a spacer and end plugs, a concept of them was established by using the basic dimension and array produced by other sub-projects. In turn, the basic design was completed by using the finite element analysis and conventional mechanical design formulae. Additionally, a welding method and equipment for a dual-cooled fuel rod specimen was also successfully developed to prepare for the irradiation tests at the HANARO. It was shown that a dual-cooed fuel for the OPR-1000 can be designed after manufacturing the partial assembly with the designed components and their drawings. The first stage was completed with passing the Gate checks proposed at the beginning. During the second stage(2010.03.∼2012.02.), researches on the mechanical behavior and structural integrity of the designed dual-cooled fuel will be conducted for preparing a license of it, which should be done when the dual-cooled fuel is commercialized

  16. High Temperature Polymer Electrolyte Fuel Cells

    DEFF Research Database (Denmark)

    Fleige, Michael

    This thesis presents the development and application of electrochemical half-cell setups to study the catalytic reactions taking place in High Temperature Polymer Electrolyte Fuel Cells (HTPEM-FCs): (i) a pressurized electrochemical cell with integrated magnetically coupled rotating disk electrode...... oxidation of ethanol is in principle a promising concept to supply HTPEM-FCs with a sustainable and on large scale available fuel (ethanol from biomass). However, the intermediate temperature tests in the GDE setup show that even on Pt-based catalysts the reaction rates become first significant...... at potentials, which approach the usual cathode potentials of HTPEM-FCs. Therefore, it seems that H3PO4-based fuel cells are not much suited to efficiently convert ethanol in accordance with findings in earlier research papers. Given that HTPEM-FCs can tolerate CO containing reformate gas, focusing research...

  17. Thermal Cooling Limits of Sbotaged Spent Fuel Pools

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Thomas G. Hughes; Dr. Thomas F. Lin

    2010-09-10

    To develop the understanding and predictive measures of the post “loss of water inventory” hazardous conditions as a result of the natural and/or terrorist acts to the spent fuel pool of a nuclear plant. This includes the thermal cooling limits to the spent fuel assembly (before the onset of the zircaloy ignition and combustion), and the ignition, combustion, and the subsequent propagation of zircaloy fire from one fuel assembly to others

  18. Model of cooling nuclear fuel rod in the nuclear reactor

    International Nuclear Information System (INIS)

    Lavicka, David; Polansky, Jiri

    2010-01-01

    The following topics are described: Some basic requirements for nuclear fuel rods; The VVER 1000 fuel rod; Classification of the two-phase flow in the vertical tube; Type of heat transfer crisis in the vertical tube; Experimental apparatus; Model of the nuclear fuel rod and spacers; Potential of the experimental apparatus (velocity profile measurement via PIV; thermal flow field measurement by the PLIF method; cooling graph in dependence on the fuel rod temperature; comparison of the hydrodynamic properties with respect to the design features of the spacers). (P.A.)

  19. Gas cooled fast breeder reactors using mixed carbide fuel

    International Nuclear Information System (INIS)

    Kypreos, S.

    1976-09-01

    The fast reactors being developed at the present time use mixed oxide fuel, stainless-steel cladding and liquid sodium as coolant (LMFBR). Theoretical and experimental designing work has also been done in the field of gas-cooled fast breeder reactors. The more advanced carbide fuel offers greater potential for developing fuel systems with doubling times in the range of ten years. The thermohydraulic and physics performance of a GCFR utilising this fuel is assessed. One question to be answered is whether helium is an efficient coolant to be coupled with the carbide fuel while preserving its superior neutronic performance. Also, an assessment of the fuel cycle cost in comparison to oxide fuel is presented. (Auth.)

  20. Heat removal in gas-cooled fuel rod clusters

    International Nuclear Information System (INIS)

    Rehme, K.

    1975-01-01

    For a thermo- and fluid-dynamic analysis of fuel rod cluster subchannels for gas-cooled breeder reactors, the following values must be verified: a) friction coefficient as flow parameter; b) Stanton number as heat transfer parameter; c) influence of spacers on friction coefficient and Stanton number; d) heat and mass exchange between subchannels with different temperatures. These parameters are established by combining results of single experiments and of integral experiments. Mention is made of further studies to be performed in order to determine the heat removal from gas-cooled fast breeder fuel elements. (HR) [de

  1. Conceptual Design of Structural Components of a Dual Cooled Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung-Kyu; Lee, Young-Ho; Lee, Kang-Hee; Kim, Jae-Yong; Yoon, Kyung-Ho

    2008-01-15

    A dual cooled fuel, featured by an internal as well as an external coolant flow passage of a fuel rod, was suggested to enable a large-scaled power-uprate of PWR plant and launched as one of the National Nuclear R and D Projects in 2007. It is necessary to make the dual cooled fuel be compatible with an OPR-1000 system to maximize the economy. Also, the structural components of the dual cooled fuel should be designed to realize their features. To this end, a conceptual design of a spacer grid, outer and center guide tubes, and top and bottom end pieces has been carried out in the project 'Development of Design Technology for Dual Cooled Fuel Structure'. For the spacer grids, it is suggested that springs and dimples are located at or near the cross points of the straps due to a considerably narrowed rod-to-rod gap. Candidate shapes of the grids were also developed and applied for domestic patents. For the outer and center guide tubes, a dual tube like a fuel rod was suggested to make the subchannel areas around the guide tubes be similar to those around the fuel rods of enlarged diameter. It was applied for the domestic patent as well. For the top and bottom end pieces, the shape and pattern have been changed from the conventional ones reflecting the fuel rods' changes. Technical issues and method of resolution for each components were listed up for a basic design works in the following years.

  2. Control and Experimental Characterization of a Methanol Reformer for a 350 W High Temperature Polymer Electrolyte Membrane Fuel Cell system

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Kær, Søren Knudsen; Jensen, Hans-Christian Becker

    , i.e. cathode and anode gas flows and temperature by using mass flow controllers and controlled heaters. Using this system the methanol reformer is characterized in its different operating points, both steady-state but also dynamically. Methanol steam reforming is a well known process, and provides...... and burner and the behaviour of the CO concentration of the reformate gas....... the high temperature waste gas from a cathode air cooled 45 cell HTPEM fuel cell stack. The MEAs used are BASF P2100 which use phosphoric acid doped polybenzimidazole type membranes; an MEA with high CO tolerance and no complex humidity requirements. The methanol reformer used is integrated into a compact...

  3. Subchannel analysis of sodium-cooled reactor fuel assemblies with annular fuel pins

    International Nuclear Information System (INIS)

    Memmott, Matthew; Buongiorno, Jacopo; Hejzlar, Pavel

    2009-01-01

    Using a RELAP5-3D subchannel analysis model, the thermal-hydraulic behavior of sodium-cooled fuel assemblies with internally and externally cooled annular fuel rods was investigated, in an effort to enhance the economic performance of sodium-fast reactors by increasing the core power density, decreasing the core pressure drop, and extending the fuel discharge burnup. Both metal and oxide fuels at high and low conversion ratios (CR=0.25 and CR=1.00) were investigated. The externally and internally cooled annular fuel design is most beneficial when applied to the low CR core, as clad temperatures are reduced by up to 62.3degC for the oxide fuel, and up to 18.5degC for the metal fuel. This could result in a power uprates of up to ∼44% for the oxide fuel, and up to ∼43% for the metal fuel. The use of duct ribs was explored to flatten the temperature distribution at the core outlet. Subchannel analyses revealed that no fuel melting would occur in the case of complete blockage of the hot interior-annular channel for both metal and oxide fuels. Also, clad damage would not occur for the metal fuel if the power uprate is 38% or less, but would indeed occur for the oxide fuel. (author)

  4. Dry fuel store for advanced gas cooled reactor fuels

    International Nuclear Information System (INIS)

    Grant, J.S.; Boocock, P.M.; Ealing, C.J.

    1992-01-01

    This paper summarizes the fuel storage requirements in Scotland and the selection of a Dry Fuel Store of the Modular Vault Dry Store (MVDS) design developed by GEC ALSTHOM Engineering Systems Limited (GECA). A similar design of store has been selected and has been constructed in the USA by Foster Wheeler Energy Corporation in collaboration with GECA

  5. Loss of spent fuel pool cooling PRA: Model and results

    International Nuclear Information System (INIS)

    Siu, N.; Khericha, S.; Conroy, S.; Beck, S.; Blackman, H.

    1996-09-01

    This letter report documents models for quantifying the likelihood of loss of spent fuel pool cooling; models for identifying post-boiling scenarios that lead to core damage; qualitative and quantitative results generated for a selected plant that account for plant design and operational practices; a comparison of these results and those generated from earlier studies; and a review of available data on spent fuel pool accidents. The results of this study show that for a representative two-unit boiling water reactor, the annual probability of spent fuel pool boiling is 5 x 10 -5 and the annual probability of flooding associated with loss of spent fuel pool cooling scenarios is 1 x 10 -3 . Qualitative arguments are provided to show that the likelihood of core damage due to spent fuel pool boiling accidents is low for most US commercial nuclear power plants. It is also shown that, depending on the design characteristics of a given plant, the likelihood of either: (a) core damage due to spent fuel pool-associated flooding, or (b) spent fuel damage due to pool dryout, may not be negligible

  6. Behaviour of gas cooled reactor fuel under accident conditions

    International Nuclear Information System (INIS)

    1991-11-01

    The Specialists Meeting on Behaviour of Gas Cooled Reactor Fuel under Accident Conditions was convened by the International Atomic Energy Agency on the recommendation of the International Working Group on Gas Cooled Reactors. The purpose of the meeting was to provide an international forum for the review of the development status and for the discussion on the behaviour of gas cooled reactor fuel under accident conditions and to identify areas in which additional research and development are still needed and where international co-operation would be beneficial for all involved parties. The meeting was attended by 45 participants from France, Germany, Japan, Switzerland, the Union of Soviet Socialists Republics, the United Kingdom, the United States of America, CEC and the IAEA. The meeting was subdivided into five technical sessions: Summary of Current Research and Development Programmes for Fuel; Fuel Manufacture and Quality Control; Safety Requirements; Modelling of Fission Product Release - Part I and Part II; Irradiation Testing/Operational Experience with Fuel Elements; Behaviour at Depressurization, Core Heat-up, Power Transients; Water/Steam Ingress - Part I and Part II. 22 papers were presented. A separate abstract was prepared for each of these papers. At the end of the meeting a round table discussion was held on Directions for Future R and D Work and International Co-operation. Refs, figs and tabs

  7. Fuel Development For Gas-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  8. Fuel performance and fission product behaviour in gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport. Refs, figs, tabs.

  9. Fuel performance and fission product behaviour in gas cooled reactors

    International Nuclear Information System (INIS)

    1997-11-01

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport

  10. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

    2012-12-01

    This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

  11. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

    2012-06-01

    This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

  12. Reprocessing technology of liquid metal cooled fast breeder reactor fuel

    International Nuclear Information System (INIS)

    Baetsle, L.H.; Broothaerts, J.; Heylen, P.R.; Eschrich, H.; Geel, J. van

    1974-11-01

    All the important aspects of LMFBR fuel reprocessing are critically reviewed in this report. Storage and transportation techniques using sodium, inert gas, lead, molten salts and organic coolants are comparatively discussed in connection with cooling time and de-activation techniques. Decladding and fuel disaggregation of UO 2 -PuO 2 fuel are reviewed according to the present state of R and D in the main nuclear powers. Strong emphasis is put on on voloxidation, mechanical pulverization and molten salt disaggregation in connection with volatilization of gaseous fission products. Release of fission gases and the resulting off-gas treatment are discussed in connection with cooling time, burn up and dissagregation techniques. The review is limited to tritium, iodine xenon-krypton and radioactive airborne particulates. Dissolution, solvent extraction and plutonium purification problems specifically connected to LMFBR fuel are reviewed with emphasis on the differences between LWR and fast fuel reprocessing. Finally the categories of wastes produced by reprocessing are analysed according to their origin in the plant and their alpha emitters content. The suitable waste treatment techniques are discussed in connection with the nature of the wastes and the ultimate disposal technique. (author)

  13. Gas-cooled Fast Reactor (GFR) fuel and In-Core Fuel Management

    International Nuclear Information System (INIS)

    Weaver, K.D.; Sterbentz, J.; Meyer, M.; Lowden, R.; Hoffman, E.; Wei, T.Y.C.

    2004-01-01

    The Gas-Cooled Fast Reactor (GCFR) has been chosen as one of six candidates for development as a Generation IV nuclear reactor based on: its ability to fully utilize fuel resources; minimize or reduce its own (and other systems) actinide inventory; produce high efficiency electricity; and the possibility to utilize high temperature process heat. Current design approaches include a high temperature (2 850 C) helium cooled reactor using a direct Brayton cycle, and a moderate temperature (550 C - 650 C) helium or supercritical carbon dioxide (S-CO 2 ) cooled reactor using direct or indirect Brayton cycles. These design choices have thermal efficiencies that approach 45% to 50%, and have turbomachinery sizes that are much more compact compared to steam plants. However, there are challenges associated with the GCFR, which are the focus of current research. This includes safety system design for decay heat removal, development of high temperature/high fluence fuels and materials, and development of fuel cycle strategies. The work presented here focuses on the fuel and preliminary in-core fuel management, where advanced ceramic-ceramic (cercer) dispersion fuels are the main focus, and average burnups to 266 M Wd/kg appear achievable for the reference Si C/(U,TRU)C block/plate fuel. Solid solution (pellet) fuel in composite ceramic clad (Si C/Si C) is also being considered, but remains as a backup due to cladding fabrication challenges, and high centerline temperatures in the fuel. (Author)

  14. Fuel assembly for gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Yellowlees, J.M.

    1976-01-01

    A fuel assembly is described for gas-cooled nuclear reactor which consists of a wrapper tube within which are positioned a number of spaced apart beds in a stack, with each bed containing spherical coated particles of fuel; each of the beds has a perforated top and bottom plate; gaseous coolant passes successively through each of the beds; through each of the beds also passes a bypass tube; part of the gas travels through the bed and part passes through the bypass tube; the gas coolant which passes through both the bed and the bypass tube mixes in the space on the outlet side of the bed before entering the next bed

  15. High Temperature Gas Cooled Reactor Fuels and Materials

    International Nuclear Information System (INIS)

    2010-03-01

    At the third annual meeting of the technical working group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), held in Vienna, in 2004, it was suggested 'to develop manuals/handbooks and best practice documents for use in training and education in coated particle fuel technology' in the IAEA's Programme for the year 2006-2007. In the context of supporting interested Member States, the activity to develop a handbook for use in the 'education and training' of a new generation of scientists and engineers on coated particle fuel technology was undertaken. To make aware of the role of nuclear science education and training in all Member States to enhance their capacity to develop innovative technologies for sustainable nuclear energy is of paramount importance to the IAEA Significant efforts are underway in several Member States to develop high temperature gas cooled reactors (HTGR) based on either pebble bed or prismatic designs. All these reactors are primarily fuelled by TRISO (tri iso-structural) coated particles. The aim however is to build future nuclear fuel cycles in concert with the aim of the Generation IV International Forum and includes nuclear reactor applications for process heat, hydrogen production and electricity generation. Moreover, developmental work is ongoing and focuses on the burning of weapon-grade plutonium including civil plutonium and other transuranic elements using the 'deep-burn concept' or 'inert matrix fuels', especially in HTGR systems in the form of coated particle fuels. The document will serve as the primary resource materials for 'education and training' in the area of advanced fuels forming the building blocks for future development in the interested Member States. This document broadly covers several aspects of coated particle fuel technology, namely: manufacture of coated particles, compacts and elements; design-basis; quality assurance/quality control and characterization techniques; fuel irradiations; fuel

  16. Comparison of fuel assemblies in lead cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Perez, A.; Sanchez, H.; Aguilar, L.; Espinosa P, G., E-mail: alejandria.peval@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico)

    2016-09-15

    This paper presents a comparison of the thermal-fluid processes in the core, fuel heat transfer, and thermal power between two fuel assemblies: square and hexagonal, in a lead-cooled fast reactor (Lfr). A multi-physics reduced order model for the analysis of Lfr single channel is developed in this work. The work focused on a coupling between process of neutron kinetic, fuel heat transfer process and thermal-fluid, in a single channel. The thermal power is obtained from neutron point kinetics model, considering a non-uniform power distribution. The analysis of the processes of thermal-fluid considers thermal expansion effects. The transient heat transfer in fuel is carried out in an annular geometry, and one-dimensional in radial direction for each axial node. The results presented in comparing these assemblies consider the temperature field in the fuel, in the thermal fluid and under steady state, and transient conditions. Transients consider flow of coolant and inlet temperature of coolant. The mathematical model of Lfr considers three main modules: the heat transfer in the annular fuel, the power generation with feedback effects on neutronic, and the thermal-fluid in the single channel. The modeling of nuclear reactors in general, the coupling is crucial by the feedback between the neutron processes with fuel heat transfer, and thermo-fluid, where is very common the numerical instabilities, after all it has to refine the model to achieve the design data. In this work is considered as a reference the ELSY reactor for the heat transfer analysis in the fuel and pure lead properties for analyzing the thermal-fluid. The results found shows that the hexagonal array has highest temperature in the fuel, respect to square array. (Author)

  17. Comparison of fuel assemblies in lead cooled fast reactors

    International Nuclear Information System (INIS)

    Perez, A.; Sanchez, H.; Aguilar, L.; Espinosa P, G.

    2016-09-01

    This paper presents a comparison of the thermal-fluid processes in the core, fuel heat transfer, and thermal power between two fuel assemblies: square and hexagonal, in a lead-cooled fast reactor (Lfr). A multi-physics reduced order model for the analysis of Lfr single channel is developed in this work. The work focused on a coupling between process of neutron kinetic, fuel heat transfer process and thermal-fluid, in a single channel. The thermal power is obtained from neutron point kinetics model, considering a non-uniform power distribution. The analysis of the processes of thermal-fluid considers thermal expansion effects. The transient heat transfer in fuel is carried out in an annular geometry, and one-dimensional in radial direction for each axial node. The results presented in comparing these assemblies consider the temperature field in the fuel, in the thermal fluid and under steady state, and transient conditions. Transients consider flow of coolant and inlet temperature of coolant. The mathematical model of Lfr considers three main modules: the heat transfer in the annular fuel, the power generation with feedback effects on neutronic, and the thermal-fluid in the single channel. The modeling of nuclear reactors in general, the coupling is crucial by the feedback between the neutron processes with fuel heat transfer, and thermo-fluid, where is very common the numerical instabilities, after all it has to refine the model to achieve the design data. In this work is considered as a reference the ELSY reactor for the heat transfer analysis in the fuel and pure lead properties for analyzing the thermal-fluid. The results found shows that the hexagonal array has highest temperature in the fuel, respect to square array. (Author)

  18. A cooling concept of spent fuels in lag storage system

    International Nuclear Information System (INIS)

    Park, Jeong-Hwa; Yoo, Jae-Hyung; Park, Hyun-Soo

    1991-01-01

    A cooling concept of spent fuels by natural convection of hot cell air in storage pits was developed. Each storage pit was considered to be located below the hot cell floor and to accommodate only one spent fuel assembly. The aim of this study is to apply an appropriate cooling system to the design of a hot cell where considerable heat-generating fuels are handled. In such operations as disassembling, rod consolidation and packaging of spent fuels, a number of assemblies are on stand-by in the cell before and/or after the operations. A lag storage system can be used for temporary storage of spent fuels in nuclear facilities. Since the air in contact with bare fuel assemblies is potentially contaminated, it must be exhausted through high-efficiency particulate air (HEPA) filters. If the storage pit is completely isolated from the hot cell space, then it will require another separate ventilation system by forced convection of air, which will result in additional cost for the construction. In this work, however, a cooling system was proposed where natural convection of hot cell air itself is achieved by thermo-syphon. The cold air from the hot cell is supplied to the inlet provided at the bottom of each pit through the gap between the concrete pit wall and the interior thermal shield. This thermal shield is needed to form flow channels for cold and heated air, and to prevent the concrete from over-heating. The heated air exhausts from the outlet located at the top of cell wall. No additional HEPA filters are needed in this system because the heated air is routed back to the hot cell due to buoyancy-induced flow. The technical feasibility of this concept was validated by thermal analyses. As the key design constraints are the surface temperature of fuel cladding and the concrete temperature of the storage pit, the thermal analyses were focused on these parameters whether they follow within allowable limits or not. (author)

  19. Fuel rod for liquid metal-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Vinz, P.

    1976-01-01

    In fuel rods for nuclear reactors with liquid-metal cooling (sodium), with stainless steel tubes with a nitrated surface as canning, superheating or boiling delay should be avoided. The inner wall of the can is provided along its total length with a helical fin of stainless steel wire (diameter 0.05 to 0.5 mm) to be wetted by hot sodium. This fin is mounted under prestressing and has a distance in winding of 1/10 of the wire diameter. (UWI) [de

  20. Cooling flow measurement in fuel elements of the RA-6

    International Nuclear Information System (INIS)

    Brollo, F; Silin, N

    2009-01-01

    Under the UBERA6 project for the core change and power increase of the RA-6 reactor, the total coolant flow was increased to meet the requirements imposed by the new operating conditions. The flow through the fuel elements is an important parameter and is difficult to determine due to the geometric complexity of the core. To ensure safe operation of the reactor, adequate safety margins must be kept for all operating conditions. In the present work we performed the direct measurement of the cooling flow rate of a fuel in the reactor core, for which we used a turbine flowmeter built specifically for this use. This helped to confirm previous results obtained during the launch, made by an indirect method based on measuring the pressure difference of the core. The turbine flowmeter was chosen due to its robustness, ease of operation and low disturbance of the input stream to the fuel. We describe the calibration of this instrument and the results of flow measurements made on some of the RA6 reactor fuel elements under conditions of zero power. [es

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

    International Nuclear Information System (INIS)

    Lee, Chi Young; Shin, Chang Hwan; Park, Ju Yong; Oh, Dong Seok; Chun, Tae Hyun; In, Wang Kee

    2012-01-01

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

  2. Visualization test facility of nuclear fuel rod emergency cooling system

    International Nuclear Information System (INIS)

    Candido, Marcos Antonio; Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Santos, Andre Augusto Campagnole

    2013-01-01

    The nuclear reactors safety is determined according to their protection against the consequences that may result from postulated accidents. The Loss of Coolant Accident (LOCA) is one the most important design basis accidents (DBA). The failure may be due to rupture of the primary loop piping. Another accident postulated is due to lack of power in the pump motors in the primary circuit. In both cases the reactor shut down automatically due to the decrease of reactivity to maintain the fissions, and to the drop of control rods. In the event of an accident it is necessary to maintain the coolant flow to remove the fuel elements residual heat, which remains after shut down. This heat is a significant amount of the maximum thermal power generated in normal operation (about 7%). Recently this event has been quite prominent in the press due to the reactor accident in Fukushima nuclear power station. This paper presents the experimental facility under rebuilding at the Thermal Hydraulic Laboratory of the Nuclear Technology Development Center (CDTN) that has the objective of monitoring and visualization of the process of emergency cooling of a nuclear fuel rod simulator, heated by Joule effect. The system will help the comprehension of the heat transfer process during reflooding after a loss of coolant accident in the fuel of light water reactor core. (author)

  3. Coated particle fuel for high temperature gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl; Nabielek, Heinz [Research Center Julich (FZJ), Julich (Germany); Kendall, James M. [Global Virtual L1c, Prescott (United States)

    2007-10-15

    applications at 850-900 .deg. C and for process heat/hydrogen generation applications with 950 .deg. C outlet temperatures. There is a clear set of standards for modern high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a 500 {mu}m diameter UO{sub 2} kernel of 10% enrichment is surrounded by a 100 {mu}m thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of 35 {mu}m thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum 1600 .deg. C afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modern coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond 1600 .deg. C for a short period of time. This work should proceed at both national and international level.

  4. Coated particle fuel for high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    Verfondern, Karl; Nabielek, Heinz; Kendall, James M.

    2007-01-01

    and for process heat/hydrogen generation applications with 950 .deg. C outlet temperatures. There is a clear set of standards for modern high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a 500 μm diameter UO 2 kernel of 10% enrichment is surrounded by a 100 μm thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of 35 μm thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum 1600 .deg. C afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modern coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond 1600 .deg. C for a short period of time. This work should proceed at both national and international level

  5. Conceptual design of reactor TRIGA PUSPATI (RTP) spent fuel pool cooling system

    International Nuclear Information System (INIS)

    Tonny Lanyau; Mazleha Maskin; Mohd Fazli Zakaria; Mohmammad Suhaimi Kassim; Ahmad Nabil Abdul Rahim; Phongsakorn Prak Tom; Mohd Fairus Abdul Farid; Mohd Huzair Hussain

    2012-01-01

    After undergo about 30 years of safe operation, Reactor TRIGA PUSPATI (RTP) was planned to be upgraded to ensure continuous operation at optimum safety condition. In the meantime, upgrading is essential to get higher flux to diversify the reactor utilization. Spent fuel pool is needed for temporary storage of the irradiated fuel before sending it back to original country for reprocessing, reuse after the upgrading accomplished or final disposal. The irradiated fuel elements need to be secure physically with continuous cooling to ensure the safety of the fuels itself. The decay heat probably still exist even though the fuel elements not in the reactor core. Therefore, appropriate cooling is required to remove the heat produced by decay of the fission product in the irradiated fuel element. The design of spent fuel pool cooling system (SFPCS) was come to mind in order to provide the sufficient cooling to the irradiated fuel elements and also as a shielding. The spent fuel pool cooling system generally equipped with pumps, heat exchanger, water storage tank, valve and piping. The design of the system is based on criteria of the primary cooling system. This paper provides the conceptual design of the spent fuel cooling system. (author)

  6. Fuel rod quenching with oxidation and precursory cooling

    International Nuclear Information System (INIS)

    Davidi, A.; Elias, E.; Olek, S.

    1999-01-01

    During a loss-of-coolant-accident in LWR fuel rods may be temporarily exposed thus reaching high temperature levels. The injection of cold water into the core, while providing the necessary cooling to prevent melting may also generate steam inducing exothermal oxidation of the cladding. A number of high temperature quenching experiments [I] have demonstrated that during the early phase of the quenching process, the rate of hydrogen generation increased markedly and the surface temperatures rose rapidly. These effects are believed to result from thermal stresses breaking up the oxide layer on the zircalloy cladding, thus exposing the inner surface to oxidizing atmosphere. Steam reacts exothermally with the metallic components of the newly formed surface causing temporarily local temperature escalation. The main objective of this study is to develop and assess a one-dimensional time-dependent rewetting model to address the problem of quenching of hot surfaces undergoing exothermic oxidation reactions. Addressing a time-dependent problem is an important aspect of the work since it is believed that the progression of a quench-front along a hot oxidizing surface is an unsteady process. Several studies dealing with time-dependent rewetting problems have been published, e.g. [2]-[5], but none considers oxidation reactions downstream of the quench-front. The main difficulty in solving time-dependent rewetting problems stems from the fact that either the quench-front velocity or the quench-front positions constitute a time-dependent eigenvalue of the problem. The model is applied to describe the interrelated processes of cooling and exothermic steam-metal reactions at the vapor zirconium-cladding interface during quenching of degraded fuel rods. A constant heat transfer coefficient is assumed upstream of the quenching front whereas the combined effect of oxidation and post dry-out cooling is described by prescribing a heat flux distribution of general form downstream. The

  7. Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

    This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

  8. Spent fuel pool cooling system upgrade for Kori Unit 1

    International Nuclear Information System (INIS)

    Sun Park, Jong; In Shin, Kyung

    2014-01-01

    Following Fukushima nuclear power plant accident, the needs for reliable performance of its own safety functions of Spent Fuel Pool Cooling System (SFPCS) has risen significantly to maintain the plant in a safe condition. Regulatory Guide 1.13 of United States Nuclear Regulatory Commission (USNRC) requires the SFPCS shall be designed safety related as Quality Group C and Seismic Category 1. However, the existing Spent Fuel Pool (SFP) of KORI Unit 1 was not designed as a safety system. In order to comply with the above licensing requirement for the extended operational life of KORI Unit 1, it has been decided to add a safety related Seismic Category 1 Makeup System to KORI Unit 1 and the existing SFPCS to be modified in dedicated channels with safety related equipment to enhance system's reliability as a means of providing diversity. This paper focuses on describing the relevant design requirements, applications, and supplemental facilities to the SFPCS of KORI Unit 1. (authors)

  9. Preliminary Study on the Fretting Wear Behaviors of a Duel Cooled Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y.H.; Lee, K.H.; Kim, H.K. [KAERI, 150 Dukjin-dong Yuseon-gu Daejeon, 305-353 (Korea, Republic of)

    2009-06-15

    Based on MIT's concept, an innovative fuel development project was launched by KAERI that a substantial power up-rating could be realized by introducing an internally and externally double cooled annular fuel for current PWR reactors. In order to apply this duel cooled fuel to an OPR 1000 reactor system, geometrical features of structural parts in a fuel assembly should be changed except an overall dimension of a fuel assembly. Typical changes are summarized as fuel rod diameter and weight, shape and position of a spacer grid spring, etc. When considering a duel cooled fuel rod, its vibration characteristic and fretting behavior should be verified because the modified shape and dimension of spacer grid spring, fuel rod diameter and weight, number of spacer grid assembly are closely related to a flow-induced vibration in a duel cooled fuel assembly. In this study, based on FIV test results of 4x4 fuel assembly, fretting wear tests of an outer duel cooled fuel rod were performed by using an embossing type spacer grid spring that could adjust its spring stiffness. The discussion was focused on the evaluation of the optimized spring stiffness and spring position in 1x1 cell by analyzing the fretting wear results. (authors)

  10. Analysis on small long life reactor using thorium fuel for water cooled and metal cooled reactor types

    International Nuclear Information System (INIS)

    Permana, Sidik

    2009-01-01

    Long-life reactor operation can be adopted for some special purposes which have been proposed by IAEA as the small and medium reactor (SMR) program. Thermal reactor and fast reactor types can be used for SMR and in addition to that program the utilization of thorium fuel as one of the candidate as a 'partner' fuel with uranium fuel which can be considered for optimizing the nuclear fuel utilization as well as recycling spent fuel. Fissile U-233 as the main fissile material for thorium fuel shows higher eta-value for wider energy range compared with other fissile materials of U-235 and Pu-239. However, it less than Pu-239 for fast energy region, but it still shows high eta-value. This eta-value gives the reactor has higher capability for obtaining breeding condition or high conversion capability. In the present study, the comparative analysis on small long life reactor fueled by thorium for different reactor types (water cooled and metal cooled reactor types). Light water and heavy water have been used as representative of water-cooled reactor types, and for liquid metal-cooled reactor types, sodium-cooled and lead-bismuth-cooled have been adopted. Core blanket arrangement as general design configuration, has been adopted which consist of inner blanket region fueled by thorium oxide, and two core regions (inner and out regions) fueled by fissile U-233 and thorium oxide with different percentages of fissile content. SRAC-CITATION and JENDL-33 have been used as core optimization analysis and nuclear data library for this analysis. Reactor operation time can reaches more than 10 years operation without refueling and shuffling for different reactor types and several power outputs. As can be expected, liquid metal cooled reactor types can be used more effective for obtaining long life reactor with higher burnup, higher power density, higher breeding capability and lower excess reactivity compared with water-cooled reactors. Water cooled obtains long life core operation

  11. Improvements in or relating to cooling systems for nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Ljubivy, A.G.; Batjukov, V.I.; Shkhian, T.G.; Fadeev, A.I.

    1980-01-01

    A cooling system is proposed which can be used to cool a set of nuclear fuel assemblies arranged in a reactor core or placed in a container for spent fuel assemblies. The object of the invention is to provide a system which would prevent leakage of coolant from the vessel in the event of a rupture of the coolant supply pipeline externally of the vessel. In the case of the reactor cooling system the level of the coolant is stopped from dropping below the level of the active portion of the fuel assemblies and thus prevents a breakdown of the reactor. (UK)

  12. Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

    International Nuclear Information System (INIS)

    Park, Jee Won; Jeong, C. J.; Yang, M. S.

    2003-03-01

    Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs

  13. Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jee Won; Jeong, C. J.; Yang, M. S

    2003-03-01

    Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs.

  14. Geometric size optimization and behavior analysis of a dual-cooled annular fuel

    International Nuclear Information System (INIS)

    Deng Yangbin; Wu Yingwei; Zhang Dalin; Tian Wenxi; Qiu Suizheng; Su Guanghui; Zhang Weixu; Wu Junmei

    2014-01-01

    The dual-cooled annular fuel is one of the innovative fuel concepts, which allows substantial power density increase while maintaining safety margins comparing with that used in currently operating PWRs. In this study, a thermal-hydraulic calculation code, on the basis of inner and outer cooling balance theory, was independently developed to optimize the geometric size of dual-cooled annular fuel elements. The optimization results show that the fuel element with the optimal geometric sizes presents fantastic symmetry in temperature distribution. The optimized geometric sizes agree well with the sizes obtained by MIT (Massachusetts Institute of Technology), which on the other side validates the code reliability and accuracy as well. In addition, a thermo-mechanical-burnup coupling code was developed to study the thermodynamic and mechanical characteristics of fuel elements with considering the irradiation and burnup effects. This coupling program was applied to perform the behavior analysis of annular fuels. The calculation results show that, when the power density increases on the order of up to 50%, the dual-cooled annular fuel elements have much lower fuel temperature and much less fission gas release comparing with conventional fuel rods. Furthermore, the results indicate that the thicknesses of inner and outer gas gap cannot remain the same with the burnup increasing due to the mechanical deformations of fuel pellets and claddings, which results in significantly asymmetric temperature distribution especially at the last phase of burnup. (author)

  15. Thermal hydraulic analysis of gas-cooled reactors with annular fuel rods

    International Nuclear Information System (INIS)

    Han, Kyu Hyun; Chang, Soon Heung

    2005-01-01

    More than half of the world's energy is used in industrial processes and for heating applications which have hardly been touched by the nuclear industry. Nuclear power could be brought into a wide range of applications for industrial processes, provided that gas outlet temperatures of gascooled reactors are sufficiently high. The most limiting core design requirement which controls the core outlet temperature is the maximum acceptable fuel compact temperature. An innovative fuel design is required for a significant decrease in the fuel temperature. This study investigated the possibilities of implementing internally and externally cooled annular fuel rods in a gas-cooled reactor

  16. Application of IAEA seals to spent-fuel cooling ponds at the Bruno Leuschner NPP, Greifswald

    International Nuclear Information System (INIS)

    Burmester, M.; Kahnmeyer, W.; Heidenreich, D; Kannenberg, D.

    1985-01-01

    A description is given of the technical appliances developed and routinely used at the Greifswald NPP to allow IAEA seals to be attached to the spent fuel cooling ponds as a whole or to their lower storage levels. (author)

  17. Operating experience feedback report: Assessment of spent fuel cooling. Volume 12

    International Nuclear Information System (INIS)

    Ibarra, J.G.; Jones, W.R.; Lanik, G.F.; Ornstein, H.L.; Pullani, S.V.

    1997-02-01

    This report documents the results of an independent assessment by a team from the Office of Analysis and Evaluation of Operational Data of spent-fuel-pool (SFP) cooling in operating nuclear power plants. The team assessed the likelihood and consequences of an extended loss of SFP cooling and suggested corrective actions, based on their findings

  18. Fossil fuel and biomass burning effect on climate - heating or cooling

    Energy Technology Data Exchange (ETDEWEB)

    Kaufman, Y.J.; Fraser, R.S.; Mahoney, R.L. (NASA/Goddard Space Flight Center, Greenbelt, MD (USA))

    1991-06-01

    Emission from burning of fossil fuels and biomass (associated with deforestation) generates a radiative forcing on the atmosphere and a possible climate change. Emitted trace gases heat the atmosphere through their greenhouse effect, while particulates formed from emitted SO{sub 2} cause cooling by increasing cloud albedos through alteration of droplet size distributions. This paper reviews the characteristics of the cooling effect and applies Twomey's theory to check whether the radiative balance favours heating or cooling for the cases of fossil fuel and biomass burning. It is also shown that although coal and oil emit 120 times as many CO{sub 2} molecules as SO{sub 2} molecules, each SO{sub 2} molecule is 50-1100 times more effective in cooling the atmosphere (through the effect of aerosol particles on cloud albedo) than a CO{sub 2} molecule is in heating it. Note that this ratio accounts for the large difference in the aerosol (3-10 days) and CO{sub 2} (7-100 years) lifetimes. It is concluded, that the cooling effect from coal and oil burning may presently range from 0.4 to 8 times the heating effect. Within this large uncertainty, it is presently more likely that fossil fuel burning causes cooling of the atmosphere rather than heating. Biomass burning associated with deforestation, on the other hand, is more likely to cause heating of the atmosphere than cooling since its aerosol cooling effect is only half that from fossil fuel burning and its heating effect is twice as large. Future increases in coal and oil burning, and the resultant increase in concentration of cloud condensation nuclei, may saturate the cooling effect, allowing the heating effect to dominate. For a doubling in the CO{sub 2} concentration due to fossil fuel burning, the cooling effect is expected to be 0.1 to 0.3 of the heating effect. 75 refs., 8 tabs.

  19. The analysis of the RA reactor irradiated fuel cooling in the spent fuel pool; Analiza hladjenja ozracenog goriva u bazenu za odlaganje reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Vrhovac, M; Afgan, N; Spasojevic, D; Jovic, V [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1985-07-01

    According to the RA reactor exploitation plan the great quantity of the irradiated spent fuel will be disposed in the reactor spent fuel pool after each reactor campaign which will including the present spent fuel inventory increase the residual power level in the pool and will soon cause the pool capacity shortage. To enable the analysis of the irradiated fuel cooling the pool and characteristic spent fuel canister temperature distribution at the residual power maximum was done. The results obtained under the various spent fuel cooling conditions in the pit indicate the normal spent fuel thermal load even in the most inconvenient cooling conditions. (author)

  20. Power sources involving ~ 300W PEMFC fuel cell stacks cooled by different media

    Directory of Open Access Journals (Sweden)

    Dudek Magdalena

    2017-01-01

    Full Text Available Two constructions of ~300W PEMFC stacks, cooled by different media, were analysed. An open-cathode ~300W PEMFC stack cooled by air (Horizon, Singapore and a PEMFC F-42 stack cooled by a liquid medium (Schunk, Germany were chosen for all of the investigations described in this paper. The potential for the design and construction of power sources involving fuel cells, as well as of a hybrid system (fuel cell-lithium battery for mobile and stationary applications, is presented and discussed. The impact of certain experimental parameters on PEMFC stack performance is analysed and discussed.

  1. Inspection of fuel elements in the cooling pond of a research reactor

    International Nuclear Information System (INIS)

    Pavlov, S.V.; Mestnikov, A.V.

    1992-01-01

    Nondestructive testing methods for fuel bundles and fuel elements in the cooling ponds of atomic power plants, using special inspection stands, have come into widespread use during the past decade. This paper describes a methodological stand that was built for the laboratory development of methods and individual units of inspection stands for fuel bundles of RBMK and VVER-1000 reactors. A complex of equipment was developed for the study of irradiated fuel elements, thus creating a methodological base for developing techniques for nondestructive testing of irradiated fuel elements and equipment to obtain information about the state of the fuel elements in a reactor expeditiously. The time required to inspect a fuel element can be shortened using some techniques simultaneously. The length of a fuel element can be measured simultaneously with visual inspection, eddy-current flaw detection can be preformed at the same time as the tranverse size of the fuel element is being determined. 6 refs., 5 figs

  2. Water cooling system for sintering furnaces of nuclear fuel pellets

    International Nuclear Information System (INIS)

    1996-01-01

    This work has as a main objective to develop a continuous cooling water system, which is necessary for the cooling of the sintering furnaces. This system is used to protect them as well as for reducing the water consumption, ejecting the heat generated into this furnaces and scattering it into the atmosphere in a fast and continuous way. The problem was defined and the reference parameters established, making the adequate research. The materials were selected as well as the length of the pipeline which will carry the secondary refrigerant fluid (water). Three possible solutions were tried,and evaluated, and from these, the thermal and economically most efficient option was selected. The layout of the solution was established and the theoretical construction of a cooling system for liquids using dichlorofluoromethane (R-22), as a refrigerant and a air cooled condenser, was accomplished. (Author)

  3. The passive safety characteristics of modular high temperature gas-cooled reactor fuel elements

    International Nuclear Information System (INIS)

    Goodin, D.T.; Kania, M.J.; Nabielek, H.; Schenk, W.; Verfondern, K.

    1988-01-01

    High-Temperature Gas-Cooled Reactors (HTGR) in both the US and West Germany use an all-ceramic, coated fuel particle to retain fission products. Data from irradiation, postirradiation examinations and postirradiation heating experiments are used to study the performance capabilities of the fuel particles. The experimental results from fission product release tests with HTGR fuel are discussed. These data are used for development of predictive fuel performance models for purposes of design, licensing, and risk analyses. During off normal events, where temperatures may reach up to 1600/degree/C, the data show that no significant radionuclide releases from the fuel will occur

  4. Supercritical water-cooled reactor fuel management and economic comparison and analysis

    International Nuclear Information System (INIS)

    Cai Guangming; Ruan Liangcheng; Liu Xuechun

    2014-01-01

    The supercritical water-cooled reactor (SCWR) is expected to have an excellent fuel economical efficiency because of its high thermal efficiency. This article compares CSR1OOO with the current mainstream PWR and ABWR on the aspect of the economical efficiency of fuel management, and finally makes an unexpected conclusion that the SCWR has worse fuel economy than others. And it remains to be deliberated whether the SCWR will be the fourth generation of nuclear system. (authors)

  5. Fuel assembly for light-water cooled nuclear reactors

    International Nuclear Information System (INIS)

    Leroux, J.C.; Burfin, P.

    1995-01-01

    In order to make easier the replacement of damaged fuel rods, a fuel assembly has been designed with a cluster of parallel fuel rods maintained in guide tubes with braces and sockets fixed on each tube ends; at least one of the fixing sockets of each tube is dismountable as well as an adapter plate on the socket, in order to lock or un-lock the guide tubes from the sockets. 11 fig

  6. Hybrid energy fuel cell based system for household applications in a Mediterranean climate

    International Nuclear Information System (INIS)

    Nižetić, S.; Tolj, I.; Papadopoulos, A.M.

    2015-01-01

    Highlights: • A hybrid energy system was proposed, combining a HT-PEM fuel cell system and a standard split heat pump system with heat recovery for household applications. • The hybrid energy system is able to produce both high and low temperature heat, electricity and cooling capacity. • The system showed high overall energy efficiency and a favorable environmental aspect. • The calculated cost of overall produced energy proved to be competitive in comparison with the average cost of electricity for households. - Abstract: In this paper, a specific hybrid energy system was proposed for household applications. The hybrid energy system was assembled from a HT-PEM fuel cell stack supplied by hydrogen via a steam reformer, where finally the majority of produced electricity is used to drive a modified split heat pump system with heat recovery (that is enabled via standard modified accumulation boilers). The system is able to produce both high and low temperature heat output (in the form of hot water), cooling thermal output and electricity. Performance analysis was conducted and the specific hybrid energy system showed high value for overall energy efficiency, for the specific case examined it reached 250%. Levelized Cost of Energy (LCOE) analysis was also carried out and the proposed hybrid energy system’s cost is expected to be between 0.09 €/kW h and 0.16 €/kW h, which is certainly competitive with the current retail electricity price for households on the EU market. Additionally, the system also has environmental benefits in relation to reduced CO 2 emissions, as estimated CO 2 emissions from the proposed hybrid energy system are expected to be at around 9.0 gCO 2 /kW h or 2.6 times less than the emissions released from the utilization of grid electricity.

  7. Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jin Ha; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O{sub 2} and (U,TRU)O{sub 2} which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O{sub 2}, (Th,Pu)O{sub 2} and (Th,TRU)O{sub 2}, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

  8. Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Choi, Jin Ha; Kim, Myung Hyun

    2016-01-01

    The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O_2 and (U,TRU)O_2 which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O_2, (Th,Pu)O_2 and (Th,TRU)O_2, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

  9. Advanced Fuel Pellet Materials and Fuel Rod Design for Water Cooled Reactors. Proceedings of a Technical Committee Meeting

    International Nuclear Information System (INIS)

    2010-10-01

    The economics of current nuclear power plants have improved through increased fuel burnup and longer fuel cycles, i.e. increasing the effective time that fuel remains in the reactor core and the amount of energy it generates. Efficient consumption of fissile material in the fuel element before it is discharged from the reactor means that less fuel is required over the reactor's life cycle, which results in lower amounts of fresh fuel, lower spent fuel storage costs, and less waste for ultimate disposal. Better utilization of fissile nuclear materials, as well as more flexible power manoeuvring, place challenging operational demands on materials used in reactor components, and first of all, on fuel and cladding materials. It entails increased attention to measures ensuring desired in-pile fuel performance parameters that require adequate improvements in fuel material properties and fuel rod designs. These are the main reasons that motivated the IAEA Technical Working Group on Fuel Performance and Technology (TWG-FPT) to recommend the organization of a Technical Committee Meeting on Advanced Fuel Pellet Materials and Fuel Rod Designs for Power Reactors. The proposal was supported by the IAEA TWGs on Advanced Technologies for Light and Heavy Water-Cooled Reactors (TWG-LWR and TWG-HWR), and the meeting was held at the invitation of the Government of Switzerland at the Paul Scherrer Institute in Villigen, from 23 to 26 November 2009. This was the third IAEA meeting on these subjects (the first was held in 1996 in Tokyo, Japan, and the second in 2003 in Brussels, Belgium), which reflects the continuous interest in the above issues among Member States. The purpose of the meeting was to review the current status in the development of fuel pellet materials and to explore recent improvements in fuel rod designs for light and heavy water cooled power reactors. The meeting was attended by 45 specialists representing fuel vendors, nuclear utilities, research and development

  10. Prediction of the Long Term Cooling Performance for the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, H. R

    2005-12-15

    In the long term cooling phase that the emergency cooling water injection ends, the performance of the residual heat removal for the 3-pin fuel test loop has been predicted by a simplified heat transfer model. In the long term cooling phase the residual heat is 1323W for PWR fuel test mode and 1449W for CANDU fuel test mode. The each residual heat is assumed as 2% of the fission power of the test fuel used in the anticipated operational occurrence and design basis accident analyses. The each fission power used for the analyses is 105% of the rated fission power in the normal operation. In the long term cooling phase the residual heat is removed to the HANARO pool through the double pressure vessels of the in-pile test section. Saturate pooling boiling is assumed on the test fuel and condensation heat transfer is expected on the inner wall of the fuel carrier and the flow divider. Natural convection heat transfer on a heated vertical wall is also assumed on the outer wall of the outer pressure vessel. The conduction heat transfer is only considered in the gap between the double pressure vessels charged with neon gas and in the downcomer filled with coolant. The heat transfer rate between the coolant temperature of 152 .deg. C in the in-pile test section and the water temperature of 45 .deg. C in the HANARO pool is predicted as about 1666W. The 152 .deg. C is the saturate temperature of the coolant pressure predicted from the MARS code. The cooling capacity of 1666W is greater than the residual heats of 1323W and 1449W. Consequently the long term cooling performance of the 3-pin fuel test loop is sufficient for the anticipated operational occurrences and design basis accidents.

  11. Replacement of the Pumps for Fuel Channel Cooling Circuit of the Maria Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krzysztoszek, G.; Mieleszczenko, W.; Moldysz, A. [National Centre for Nuclear Research, Otwock–Świerk (Poland)

    2014-08-15

    The high flux Maria research reactor is operated by the National Centre for Nuclear Research in Świerk. It is a pool type reactor with pressurized fuel channels located in the beryllium matrix. According to the Global Threat Reduction Initiative programme our goal is to convert the Maria reactor from HEU to LEU fuel. Hydraulic losses in the new LEU fuel produced by CERCA are about 30% higher than the existing HEU fuel of type MR-6. For the MR-6 fuel were installed four two speed pumps. These pumps performed the function of the main circulations pumps during reactor operation with residual pumping power provided by emergency pumps. In the new system four main pumps will be used for circulating coolant while the reactor is operation with three auxiliary pumps for decay heat removal after reactor shutdown, meaning that the conversion of Maria research reactor will be possible after increasing flow in the primary cooling circuit of the fuel channels. The technical design of replacement of the pumps in the primary fuel channel cooling circuit was finished in April 2011 and accepted by the Safety Committee. After delivery of the new pumps we are planning to upgrade the primary fuel channel cooling circuit during October–November 2012. (author)

  12. Failed fuel identification techniques for liquid-metal cooled reactors

    International Nuclear Information System (INIS)

    Lambert, J.D.B.; Gross, K.C.; Mikaili, R.; Frank, S.M.; Cutforth, D.C.; Angelo, P.L.

    1995-01-01

    The Experimental Breeder Reactor II (EBR-II), located in Idaho and operated for the US Department of Energy by Argonne National Laboratory, has been used as an irradiation testbed for LMR fuels and components for thirty years. During this time many endurance tests have been carried out with experimental LMR metal, oxide, carbide and nitride fuel elements, in which cladding failures were intentionally allowed to occur. This paper describes methods that have been developed for the detection, identification and verification of fuel failures

  13. Detection of fuel element failures at the gas cooled reactors

    International Nuclear Information System (INIS)

    Bremsak, P.; Ranzel, S.; Ogrin, R.; Pisanski, C.

    1964-01-01

    This paper describes the system for measuring the concentration of released Kr and Xe in the cooling gas. The system developed in the Jozef Stefan Institute is efficient and suitable for application The method is based on electrostatic collection of daughter elements from radioactive decay of Xe and Kr

  14. Caramel, uranium oxide fuel plates for water cooled reactors

    International Nuclear Information System (INIS)

    Bussy, Pierre; Delafosse, Jacques; Lestiboudois, Guy; Cerles, J.-M.; Schwartz, J.-P.

    1979-01-01

    The fuel is composed of thin plates assembled parallel to each other to form bundles or assemblies. Each plate is composed of a pavement of uranium oxide pellets, insulated from each other by a zircaloy cladding. The 235 U enrichment does not exceed 8%. The range of uses for this fuel extends from electric power generating reactors to irradiation reactors for research work. A parametric study in test loops has made it possible to determine the operating limits of this thick fuel, without bursting. The resulting diagram gives the permissible power densities, with and without cycling for specific burn-ups beyond 50,000 MWd/t. The thinnest plates were also irradiated in total in the form of advance assemblies irradiated in the core of the OSIRIS pile prior to its transformation. This transformation and the operation of this reactor with a core of 'Caramel' elements is the main trial experiment of this fuel [fr

  15. Vented fuel experiment for gas-cooled fast reactor application

    International Nuclear Information System (INIS)

    Longest, A.W.; Gat, U.; Conlin, J.A.; Campana, R.J.

    1975-01-01

    A pressure-equalized and vented fuel rod is being irradiated in an instrumented capsule designated GB-10 to approximately 100 MWd/kg-heavy metal. The fuel is a sol-gel derived 88 atom-percent uranium (approximately 9 percent 235 U) 12 atom-percent plutonium oxide, and the cladding is 20 percent cold-worked 316 stainless steel. The capsule is being irradiated in the Oak Ridge Research Reactor (ORR) and has exceeded a burnup of 70 MWd/kg. The fuel has been operated at linear power rates of 39 and 44 kW/ m, and peak outer cladding temperature of 565 0 and 630 0 C respectively. A similar fuel rod in a previous capsule (GB-9) was subjected to 48 kW/m (685 0 C). 4 references. (auth)

  16. Fission product release from fuel of water-cooled reactors

    International Nuclear Information System (INIS)

    Strupczewski, A.; Marks, P.; Klisinska, M.

    1997-01-01

    The report contains a review of theoretical models and experimental works of gaseous and volatile fission products from uranium dioxide fuel. The experimental results of activity release at low burnup and the model of fission gas behaviour at initial stage of fuel operational cycle are presented. Empirical models as well as measured results of transient fission products release rate in the temperature up to UO 2 melting point, with consideration of their chemical reactions with fuel and cladding, are collected. The theoretical and experimental data were used for calculations of gaseous and volatile fission products release, especially iodine and caesium, to the gas volume of WWER-1000 and WWER-440 type fuel rods at low and high burnup and their further release from defected rods at the assumed loss-of-coolant accident. (author)

  17. Process and device for cooling of nuclear reactor fuel elements enclosed in a transport container

    International Nuclear Information System (INIS)

    Stiefel, M.

    1986-01-01

    In order to remove the post-decay heat of the fuel elements contained in them, transport containers for burnt-up fuel elements can be connected to a water cooling circuit. In order to avoid thermal shocks, a tenside forming foam and air are introduced into the cooling circuit before its entry into the transport container in the direction of flow. The tenside and air continue to be supplied until the temperature inside the transport container has fallen below the temperature at which the foam is destroyed. By adding tenside and air, a two phase mixture is produced, which foams greatly when it enters the transport container and which cools the fuel elements so as to protect them.(orig./HP) [de

  18. High-burnup/low-cooling-time fuel carrying capacity of the GA-4 and GA-9 spent fuel shipping casks

    International Nuclear Information System (INIS)

    Boshoven, J.K.; Hopf, J.E.

    1994-01-01

    In response to utilities' projected needs to ship higher burnup spent fuel, General Atomics (GA) has performed shielding and thermal analysis for the GA-4 and GA-9 legal weight shipping casks to determine the minimum cooling times for various burnup levels for fully loaded GA-4 and GA-9 casks and reduced payloads for the casks. Tables are provided in the paper which show the minimum cooling time for a given burnup and payload for each of the casks. The analyses show that the GA-4 and GA-9 casks can carry at least as many high-burnup and/or short-cooling-time spent fuel assemblies as present day shipping casks. In addition, the GA casks are able to carry at least twice as many assemblies as the present day shipping casks if the spent fuel burnup levels and/or cooling times are open-quotes coolerclose quotes or open-quotes as coolclose quotes as their design basis fuels. The increased shipping capacity for these more common open-quotes coolerclose quotes assemblies allows fewer shipments and therefore increases the efficiency and lowers predicted risks of the transport system

  19. Thermal analyses for the rack design with spent fuel pool during the loss of cooling accident

    Energy Technology Data Exchange (ETDEWEB)

    Yeh, C-L.; Chen, Y-S.; Chen, B-Y., E-mail: clinyeh@iner.gov.tw, E-mail: yschen@iner.gov.tw, E-mail: onepicemine@iner.gov.tw [Inst. of Nuclear Energy Research, Taoyuan County, Taiwan (China); Tseng, Y-S., E-mail: ystseng@mx.nthu.edu.tw [National Tsing Hua Univ., Engineering and System Science, Hsinchu, Taiwan (China); Wei, W-C., E-mail: hn150456@iner.gov.tw [Inst. of Nuclear Energy Research, Taoyuan County, Taiwan (China)

    2014-07-01

    Alternative fuel arrangements separating the latest fuels discharge from the reactor core are proposed, such as the 1x4 configuration in which the hot assembly is surrounded by 4 assemblies with much lower decay heat. For the rack design in the BWR spent fuel pool design, the lateral flow is eliminated by solid walls. In this study, cooling enhancement of splitting fuel rack is investigated using Computational Fluid Dynamics (CFD). The fuels in the pool are modeled by porous medium. Separating the fuel rack by a distance of 10 cm can lower the peak cladding temperature and the natural convection between the fuels and then earns more response time for the site people to implement necessary mitigation actions. (author)

  20. Metal fuel development and verification for prototype generation- IV Sodium- Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Cheon, Jin Sik; Kim, Sung Ho; Park, Jeong Yong; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U -transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

  1. Metal Fuel Development and Verification for Prototype Generation IV Sodium-Cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    Chan Bock Lee

    2016-10-01

    Full Text Available Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic–martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

  2. Fuel particles in the Chernobyl cooling pond: current state and prediction for remediation options

    International Nuclear Information System (INIS)

    Bulgakov, A.; Konoplev, A.; Smith, J.; Laptev, G.; Voitsekhovich, O.

    2009-01-01

    During the coming years, a management and remediation strategy for the Chernobyl cooling pond (CP) will be implemented. Remediation options include a controlled reduction in surface water level of the cooling pond and stabilisation of exposed sediments. In terrestrial soils, fuel particles deposited during the Chernobyl accident have now almost completely disintegrated. However, in the CP sediments the majority of 90 Sr activity is still in the form of fuel particles. Due to the low dissolved oxygen concentration and high pH, dissolution of fuel particles in the CP sediments is significantly slower than in soils. After the planned cessation of water pumping from the Pripyat River to the Pond, significant areas of sediments will be drained and exposed to the air. This will significantly enhance the dissolution rate and, correspondingly, the mobility and bioavailability of radionuclides will increase with time. The rate of acidification of exposed bottom sediments was predicted on the basis of acidification of similar soils after liming. Using empirical equations relating the fuel particle dissolution rate to soil and sediment pH allowed prediction of fuel particle dissolution and 90 Sr mobilisation for different remediation scenarios. It is shown that in exposed sediments, fuel particles will be almost completely dissolved in 15-25 years, while in parts of the cooling pond which remain flooded, fuel particle dissolution will take about a century

  3. Fuel element replacement and cooling water activity at the musashi reactor

    International Nuclear Information System (INIS)

    Nozaki, Tetsuya; Honda, Teruyuki; Horiuchi, Norikazu; Aizawa, Otohiko; Sato, Tadashi

    1989-01-01

    The Musashi Institute of Technology Research Reactor (TRIGA 11, 100 kW) has been operated without serious problems since 1963. However, because there is no more spare fuel element, it was necessary to decide how to solve the problem. In the end, it was decided to obtain many stainless steel-clad fuel elements and operate with those fuel elements only, under the auspices of the Ministry of Education, Science and Culture. The bulk shielding experimental pool was remodeled as the storage for spent fuel elements, where the neutrons from the thermalizing column were shielded with cadmium and boron polyethylene plates. The equipment for transferring spent fuel elements was built and temporarily set up between the core tank and the new storage. These works were started in 1983, and finished in 1985. After the reactor was restarted, the count rate of the conventional cooling water monitor which was set in the cooling system using a GM counter drastically decreased. The spent fuel storage, the equipment and the works for fuel transfer, and the radioactivity of cooling water are reported. (K.I.)

  4. Water supply method to the fuel cell cooling water system; Nenryo denchi reikyakusuikei eno kyusui hoho

    Energy Technology Data Exchange (ETDEWEB)

    Urata, T. [Tokyo (Japan); Nishida, S. [Tokyo (Japan)

    1996-12-17

    The conventional fuel cell has long cooling water piping ranging from the fuel cell exit to the steam separator; in addition, the supply water is cooler than the cooling water. When the amount of supply water increases, the temperature of the cooling water is lowered, and the pressure fluctuation in the steam separator becomes larger. This invention relates to the water supply method of opening the supply water valve and supplying water from the supply water system to the cooling water system in accordance with the signal of the level sensor of the steam separator, wherein opening and closing of the supply valve are repeated during water supply. According to the method the pressure drop in every water supply becomes negligibly small; therefore, the pressure fluctuation of the cooling water system can be made small. The interval of the supply water valve from opening to closing is preferably from 3 seconds to 2 minutes. The method is effective when equipment for recovering heat from the cooling water is installed in the downstream pipeline of the fuel cell. 2 figs.

  5. Spent fuel pool spray cooling system for the AP1000 {sup registered}

    Energy Technology Data Exchange (ETDEWEB)

    Vujic, Zoran; Sassen, Felix; Tietsch, Wolfgang [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2013-07-01

    The AP1000 {sup registered} plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for Design Basis Events and Beyond Design Basis Accidents (BDBA). The AP1000 {sup registered} plant lines of defense with respect to Spent Fuel Pool (SFP) cooling are as follows: 1. During normal and abnormal conditions, defense-in-depth and duty systems provide highly reliable SFP cooling, supplied by offsite AC power or the onsite Standby Diesel Generators. 2. For unlikely events with extended loss of AC power (i.e. station black-out) and/or loss of heat sink, spent fuel cooling can be still provided indefinitely by: 2a. Passive systems, requiring minimal or no operator actions, sufficient for at least 72 hours under all possible loading conditions. 2b. After 3 days, several different means are provided to continue SFP cooling using installed plant equipment as well as off-site equipment with built-in connections. 3. Even for BDBA with postulated SFP damage and multiple failures in the passive safety-related systems and in the defense-in-depth active systems, the AP1000 {sup registered} SFP Spray System provides an additional line of defense to prevent spent fuel damage. (orig.)

  6. A comparative study on recycling spent fuels in gas-cooled fast reactors

    International Nuclear Information System (INIS)

    Choi, Hangbok; Baxter, Alan

    2010-01-01

    This study evaluates advanced Gas-cooled Fast Reactor (GFR) fuel cycle scenarios which are based on recycling spent nuclear fuel for the sustainability of nuclear energy. A 600 MWth GFR was used for the fuel cycle analysis, and the equilibrium core was searched with different fuel-to-matrix volume ratios such as 70/30 and 60/40. Two fuel cycle scenarios, i.e., a one-tier case combining a Light Water Reactor (LWR) and a GFR, and a two-tier case using an LWR, a Very High Temperature Reactor (VHTR), and a GFR, were evaluated for mass flow and fuel cycle cost, and the results were compared to those of LWR once-through fuel cycle. The mass flow calculations showed that the natural uranium consumption can be reduced by more than 57% and 27% for the one-tier and two-tier cycles, respectively, when compared to the once-through fuel cycle. The transuranics (TRU) which pose a long-term problem in a high-level waste repository, can be significantly reduced in the multiple recycle operation of these options, resulting in more than 110 and 220 times reduction of TRU inventory to be geologically disposed for the one-tier and two-tier fuel cycles, respectively. The fuel cycle costs were estimated to be 9.4 and 8.6 USD/MWh for the one-tier fuel cycle when the GFR fuel-to-matrix volume ratio was 70/30 and 60/40, respectively. However the fuel cycle cost is reduced to 7.3 and 7.1 USD/MWh for the two-tier fuel cycle, which is even smaller than that of the once-through fuel cycle. In conclusion the GFR can provide alternative fuel cycle options to the once-through and other fast reactor fuel cycle options, by increasing the natural uranium utilization and reducing the fuel cycle cost.

  7. Concept of safe tank-type water cooled and moderated reactor with HTGR microparticle fuel compacts

    International Nuclear Information System (INIS)

    Gol'tsev, A.O.; Kukharkin, N.E.; Mosevitskij, I.S.; Ponomarev-Stepnoj, N.N.; Popov, S.V.; Udyanskij, Yu.N.; Tsibul'skij, V.F.

    1993-01-01

    Concept of safe tank-type water-cooled and moderated reactor on the basis of HTGR fuel microparticles which enable to avoid environment contamination with radioactive products under severe accidents, is proposed. Results of neutron-physical and thermal-physical studies of water cooled and moderated reactor with HTGR microparticle compacts are presented. Characteristics of two reactors with thermal power of 500 and 1500 MW are indicated within the concept frames. The reactor behaviour under severe accident connected with complete loss of water coolant is considered. It is shown that under such an accident the fission products release from fuel microparticles does not occur

  8. Development of Test Methods for Structural Components of a Dual Cooled Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Ho; Lee, Kang Hee; Kim, Hyung Kyu; Yoon, Kyung Ho; Kim, Jae Yong

    2009-01-15

    The most unique feature of a dual-cooled fuel is that the outer diameter of a fuel rod is considerably increased due to an internal coolant passage additionally formed inside the rod. This increases the fuel rod's weight and decreases the gap between the fuel rods. Change of the weight and gap causes the shape and the performance of a fuel rod support structure to be necessarily altered. It also alters the flow-induced vibration (FIV) as well as the fretting wear characteristics of a fuel rod. These are directly related with the integrity of the rod so that they should be investigated in the design stage. Finite element analysis and semi-empirical formulae can be used to roughly investigate the support performance and FIV of a fuel rod, respectively. However, the fretting wear characteristic can be investigated only through an experiment. The support performance and FIV need experiment as well to obtain the characteristics more accurately. Therefore, experimental investigation of those has been included in the present project scope. In the second year, it has been planned to establish the experimental devices and technologies to accommodate the altered dimensions and feature of a dual-cooled fuel. As a result, devices to obtain the characteristics of fuel rod supports and holddown springs were developed. As for the FIV and fretting wear characteristics, the existing facilities were modified. Experimental procedures were also re-established this year.

  9. Innovate pin design for Sphere-pac fuel in sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Pouchon, Manuel A.; Niceno, Bojan; Krepel, Jiri

    2011-01-01

    The paper discusses a new fuel element type, which combines a particle fuel concept, the Sphere-pac, with a new pin design which features internal cooling. Particle fuels are auspicious when considering a closed fuel cycle, where minor actinide containing fuels must be fabricated. The principle advantage lies in their production simplicity with much less maintenance intensive mechanical devices. Furthermore the Sphere-pac is usually produced by a wet and therefore powder-less route. Therefore the implementation in a remotely controlled and heavily shielded environment becomes easier to realize. Besides the advantages in the production process, the Sphere-pac bears one important disadvantage: the lower thermal conductivity of the particle arrangement, and the therefore higher peak temperatures in the fuel. Consequently a new fuel design is suggested in this paper. It offers an internal cooling channel and therefore smaller maximal fuel distances to the coolant. As the concept is new, the most important aspects are studied; these are the neutronics, the temperature profile in the fuel plus thermal-hydraulics aspects. (author)

  10. gamma-ray spectra measurements for long cooled MOX spent fuels

    International Nuclear Information System (INIS)

    Murakami, Kiyonobu; Kobayashi, Iwao

    1993-09-01

    Gamma-ray spectra of spent fuels have important informations in the estimation of burnup rate, concentration of fission products, cooling time and etc. which are required in the fuel loading control of reactors and special nuclear materials accountancy from the view point of safe guard. Although, some available data are given about uranium dioxide fuels, few data are given about uranium and plutonium dioxide mixtures (MOX fuels). Especially, there is few data about MOX fuels which are irradiated in thermal reactors and cooled more than ten years. Gamma-ray spectra are measured for PuO 2 -UO 2 fuel rods (IFA-159, IFA-160) which are irradiated at HBWR in Norway up to 9,420 and 5,340MWd/t respectively. Gamma-ray spectra had been measured about the two fuels ten years ago at the spent fuel pond of Japan Demonstration Reactor (JPDR). The objectives of this measurement is to know how decayed the gamma-ray spectra in these ten years and some fission products are there which are effective to estimate burnup rate of spent MOX fuels. (author)

  11. Retrofitting a spent fuel pool spray system for alternative cooling as a strategy for beyond design basis events

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Christoph; Vujic, Zoran [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2017-06-15

    Due to requirements for nuclear power plants to withstand beyond design basis accidents, including events such as happened in 2011 in the Fukushima Daiichi Nuclear Power Plant in Japan, alternative cooling of spent fuel is needed. Alternative spent fuel cooling can be provided by a retrofitted spent fuel pool spray system based on the AP1000 plant design. As part of Krsko Nuclear Power Plant's Safety Upgrade Program, Krsko Nuclear Power Plant decided on, and Westinghouse successfully designed a retrofit of the AP1000 {sup registered} plant spent fuel pool spray system to provide alternative spent fuel cooling.

  12. Heat diffusion in cylindrical fuel elements of water cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Randles, J [Technical Assessments and Services Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1961-09-15

    This report contains a theoretical study of heat diffusion in the cylindrical fuel elements of water reactors. After setting up appropriate boundary conditions on the temperature, the steady state Fourier equation is solved both for a flat and a tilted fission power source. It is shown that source tilting does not have an appreciable effect on the peak fuel temperature while the heat flux to the coolant suffers a circumferential variation of less than a half of that of the fission power. In the last section, the theory is extended to include the effect of a flat, time dependent fission power. The time dependent Fourier equation is solved by means of a Dini series of Bessel functions which is shown to be rapidly convergent. From this series is derived expressions for the fuel element transfer functions required in reactor servo-analysis. These have the form of a rapidly convergent series of time-lag terms. (author)

  13. Characterisation and Modelling of a High Temperature PEM Fuel Cell Stack using Electrochemical Impedance Spectroscopy

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Jespersen, Jesper Lebæk; Schaltz, Erik

    2009-01-01

    temperature PEM (HTPEM) fuel cell stack. A Labview virtual instrument has been developed to perform the signal generation and data acquisition which is needed to perform EIS. The typical output of an EIS measurement on a fuel cell, is a Nyquist plot, which shows the imaginary and real part of the impedance...

  14. Vented fuel experiment for gas-cooled fast reactor application

    International Nuclear Information System (INIS)

    Longest, A.W.; Gat, U.; Conlin, J.A.; Campana, R.J.

    1976-01-01

    A pressure-equalized and vented fuel rod is being irradiated in an instrumented capsule designated GB-10 to approximately 100MWd/kg-heavy metal. The fuel is a sol-gel-derived 88 at.% uranium (approximately 9% 235 U) and 12 at.% plutonium oxide, and the cladding is 20% cold-worked 316 stainless steel. The capsule is being irradiated in the Oak Ridge Research Reactor (ORR) and has exceeded a burnup of 70MWd/kg. The fuel has been operated at linear power rates of 39 and 44kW/m, and peak outer cladding temperature of 565 and 630 0 C respectively. A similar fuel rod in a previous capsule (GB-9) was subjected to 48kW/m (685 0 C). Helium gas sweeps through any portion of the three regions of the fuel rod, namely: fuel, blanket, and charcoal trap. The charcoal trap is operated at about 300 0 C. An on-line Ge(Li) detector is used to analyse release rates of several gamma-emitting noble gas isotopes. Analyses are performed primarily on sweep gas flowing through the entire fuel rod, and for sweeps over the top of the charcoal trap. Sweep gas samples are analyzed for stable noble gas isotopes. Results in the form of ratios of release rate over birth rate (R/B) and venting rate over birth rate (V/B) are derived. R/B rates range from 10 -4 % to 30% while V/B ranges from 10 -6 % to 30%. Flow conductance in the capsule was monitored by recording the flow rate and pressure drop across the fuel rod and inlet sweep line. The flow conductance has been falling with increasing burnup, currently restricting the flow to about 20ml (s.t.p.)/min at a pressure difference of about 1.5MPa. Venting rates of the gaseous fission products as a function of gas pressure in the range 6.9 to 1.4MPa have also been measured. Planned future experiments include the monitoring of tritium release, venting and cladding permeation rates, and its molecular form. First measurements have been made. A simulated leak experiment will determine the mixture of fission gases as a function of flow rate and the most

  15. Investigation of heating and cooling in a stand-alone high temperature PEM fuel cell system

    International Nuclear Information System (INIS)

    Zhang, Caizhi; Yu, Tao; Yi, Jun; Liu, Zhitao; Raj, Kamal Abdul Rasheedj; Xia, Lingchao; Tu, Zhengkai; Chan, Siew Hwa

    2016-01-01

    Highlights: • Heating-up and cooling-down processes of HT-PEMFC are the mainly interested topics. • Dynamic behaviours, power and energy demand of the heating and cooling was studied. • Hybrid system based on LiFeYPO_4 battery for heating and cooling is built and tested. • The concept of combining different heating sources together is recommended. - Abstract: One key issue pertaining to the cold-start of High temperature PEM fuel cell (HT-PEMFC) is the requirement of high amount of thermal energy for heating up the stack to a temperature of 120 °C or above before it can generate electricity. Furthermore, cooling down the stack to a certain temperature (e.g. 50 °C) is necessary before stopping. In this study, the dynamic behaviours, power and energy demand of a 6 kW liquid cooled HT-PEMFC stack during heating-up, operation and cooling-down were investigated experimentally. The dynamic behaviours of fuel cell under heating-up and cooling-down processes are the mainly interested topics. Then a hybridisation of HT-PEMFC with Li-ion battery to demonstrate the synergistic effect on dynamic behaviour was conducted and validated for its feasibility. At last, the concept of combining different heating sources together is analysed to reduce the heating time of the HT-PEMFC as well.

  16. Gas-cooled fast reactor fuel-cost assessment. Final report, October 1978-September 1979

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, M.L.

    1979-01-01

    This program, contracted to provide a Gas Cooled Fast Reactor (GCFR) fuel assembly fabrication cost assessment, comprised the following basic activities: establish agreement on the ground rules for cost assessment, prepare a fuel factory flow sheet, and prepare a cost assessment for fuel assembly fabrication. Two factory sizes, 250 and 25 MTHM/year, were considered for fuel assembly fabrication cost assessment. The work on this program involved utilizing GE LMFBR cost assessment and fuel factory studies experience to provide a cost assessment of GCFR fuel assembly fabrication. The recent impact of highly sensitive safety and safeguards environment policies on fuel factory containment, safety, quality assurance and safeguards costs are significantly higher than might have been expected just a few years ago. Fuel assembly fabrication costs are significant because they represent an estimated 30 to 60% of the total fuel cycle costs. In light of the relative high cost of fabrication, changes in the core and assembly design may be necessary in order to enhance the overall fuel cycle economics. Fabrication costs are based on similar operations and experience used in other fuel cycle studies. Because of extrapolation of present technology (e.g., remote fuel fabrication versus present contact fabrication) and regulatory requirements, conservative cost estimates were made.

  17. Holding device for gas-cooled reactor fuel elements

    International Nuclear Information System (INIS)

    Hensolt, T.

    1980-01-01

    The sheathed fuel elements of the GCFR are inserted with their pedestal in a grid plate arranged below the reactor core and are clamped there. The clamping force as well as the force required for hydraulic holding-down against the flow pressure of the coolant are applied through the differential pressure between inlet and outlet of the coolant. (DG) [de

  18. Optimization of the fuel assembly for the Canadian Supercritical Water-cooled Reactor (SCWR)

    Energy Technology Data Exchange (ETDEWEB)

    French, C.; Bonin, H.; Chan, P., E-mail: Corey.French@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

    2013-07-01

    A parametric optimization of the Canadian Supercritical Water-cooled Reactor (SCWR) lattice geometry and fresh fuel content is performed in this work. With the potential to improve core physics and performance, significant gains to operating and safety margins could be achieved through slight progressions. The fuel performance codes WIMS-AECL and SERPENT are used to calculate performance factors, and use them as inputs to an optimization algorithm. (author)

  19. Fuel Cooling in Absence of Forced Flow at Shutdown Condition with PHTS Partially Drained

    Energy Technology Data Exchange (ETDEWEB)

    Parasca, L.; Pecheanu, D.L., E-mail: laurentiu.parasca@cne.ro, E-mail: doru.pecheanu@cne.ro [Cernavoda Nuclear Power Plant, Cernavoda (Romania)

    2014-09-15

    During the plant outage for maintenance on primary side (e.g. for the main Heat Transport System pumps maintenance, the Steam Generators inspection), there are situations which require the primary heat transport system (HTS) drainage to a certain level for opening the circuit. The primary fuel heat sink for this configuration is provided by the shutdown cooling system (SDCS). In case of losing the forced cooling (e.g. due to the loss of SDCS, design basis earthquake-DBE), flow conditions in the reactor core may become stagnant. Inside the fuel channels, natural circulation phenomena known as Intermittent Buoyancy Induced Flow (IBIF) will initiate, providing an alternate heat sink mechanism for the fuel. However, this heat sink is effective only for a limited period of time (recall time). The recall time is defined as the elapsed time until the water temperature in the HTS headers exceeds a certain limit. Until then, compensatory measures need to be taken (e.g. by re-establishing the forced flow or initiate Emergency Core Cooling system injection) to preclude fuel failures. The present paper briefly presents the results of an analysis performed to demonstrate that fuel temperature remains within acceptable limits during IBIF transient. One of the objectives of this analysis was to determine the earliest moment since the reactor shut down when maintenance activities on the HTS can be started such that IBIF is effective in case of losing the forced circulation. The resulting peak fuel sheath and pressure tube temperatures due to fuel heat up shall be within the acceptable limits to preclude fuel defect or fuel channel defects.Thermalhydraulic circuit conditions were obtained using a CATHENA model for the primary side of HTS (drained to a certain level), an ECC system model and a system model for SDCS. A single channel model was developed in GOTHIC code for the fuel assessment analysis. (author)

  20. Assessment of the dry process fuel sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok

    2004-04-01

    The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was {approx}50% and most of the fission products were removed.

  1. Assessment of the dry process fuel sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Choi, Hang Bok

    2004-04-01

    The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was ∼50% and most of the fission products were removed

  2. Design of a spherical fuel element for a gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Van Rooijen, W.F.G.; Kloosterman, J.L.; Van Dam, H.; Van der Hagen, T.H.J.J.

    2004-01-01

    A study is undertaken to develop a fuel cycle for a gas-cooled fast reactor (GCFR). The design goals are: highly efficient use of (depleted) uranium, application of Pu recycled from LWR discharge as fissile material, high temperature output and simplicity of design. The design focuses on spherical TRISO-like fuel elements, a homogeneous core at start-up, providing for easy fuel fabrication, and self-breeding capability with a flat k eff with burn-up. Nitride fuel ( 15 N > 99%) has been selected because of its favourable thermal conductivity, high heavy metal density and compatibility with PUREX reprocessing. Two core concepts have been studied: one with coated particles embedded inside fuel pebbles, and one with coated particles cooled directly by helium. The result is that a flat k eff can be achieved for a long period of time, using coated particles cooled directly, with a homogeneous core at, start-up, with a closed fuel cycle and a simple refuelling and reprocessing scheme. (author)

  3. Fuel element replacement and cooling water radioactivity at the Musashi reactor

    International Nuclear Information System (INIS)

    Nozaki, T.; Honda, T.; Horiuchi, N.; Aizawa, O.; Sato, T.

    1988-01-01

    The Musashi reactor (TRIGA-II, 100kW) has been operated without any serious troubles since 1963. In 1985 the old Al-cladded fuel elements were replaced with new stainless cladded ones in order to insure a long and safe operation. By using a semi-automatic equipment the old fuel elements have been transferred into the bulk-shielding experimental pool, which was remodelled for the spent-fuel storage. In order to reduce the exposure during the transfer work, the old fuel elements were cooled in the core tank for 3 months. After the replacement, the radioactivities in the cooling water have been drastically changed. The activity of Na-24 decreased about one decade, and the activities of Cr-51, Mn-54, Mn-56, Co-58 and Co-60 increased about two decades. At this conference we will report on the following points: (1) semi-automatic equipment for the transportation of the Al-cladded spent fuel, (2) structure of spent-fuel storage pool, and (3) radioactivity change in the cooling water. (author)

  4. A combined gas cooled nuclear reactor and fuel cell cycle

    Science.gov (United States)

    Palmer, David J.

    Rising oil costs, global warming, national security concerns, economic concerns and escalating energy demands are forcing the engineering communities to explore methods to address these concerns. It is the intention of this thesis to offer a proposal for a novel design of a combined cycle, an advanced nuclear helium reactor/solid oxide fuel cell (SOFC) plant that will help to mitigate some of the above concerns. Moreover, the adoption of this proposal may help to reinvigorate the Nuclear Power industry while providing a practical method to foster the development of a hydrogen economy. Specifically, this thesis concentrates on the importance of the U.S. Nuclear Navy adopting this novel design for its nuclear electric vessels of the future with discussion on efficiency and thermodynamic performance characteristics related to the combined cycle. Thus, the goals and objectives are to develop an innovative combined cycle that provides a solution to the stated concerns and show that it provides superior performance. In order to show performance, it is necessary to develop a rigorous thermodynamic model and computer program to analyze the SOFC in relation with the overall cycle. A large increase in efficiency over the conventional pressurized water reactor cycle is realized. Both sides of the cycle achieve higher efficiencies at partial loads which is extremely important as most naval vessels operate at partial loads as well as the fact that traditional gas turbines operating alone have poor performance at reduced speeds. Furthermore, each side of the cycle provides important benefits to the other side. The high temperature exhaust from the overall exothermic reaction of the fuel cell provides heat for the reheater allowing for an overall increase in power on the nuclear side of the cycle. Likewise, the high temperature helium exiting the nuclear reactor provides a controllable method to stabilize the fuel cell at an optimal temperature band even during transients helping

  5. Development of an internally cooled annular fuel bundle for pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, H.; Armstrong, J.; Kittmer, A.; Zhuchkova, A.; Xu, R.; Hyland, B.; King, M.; Nava-Dominguez, A.; Livingstone, S.; Bergeron, A. [Atomic Energy of Canada, Ltd., Chalk River Laboratories, Chalk River, ON (Canada)

    2013-07-01

    A number of preliminary studies have been conducted at Atomic Energy of Canada Limited to explore the potential of using internally cooled annular fuel (ICAF) in CANDU reactors including finite element thermo-mechanical modelling, reactor physics, thermal hydraulics, fabrication and mechanical design. The most compelling argument for this design compared to the conventional solid-rod design is the significant reduction in maximum fuel temperature for equivalent LERs (linear element ratings). This feature presents the potential for power up-rating or higher burnup and a decreased defect probability due to in-core power increases. The thermal-mechanical evaluation confirmed the significant reduction in maximum fuel temperatures for ICAF fuel compared to solid-rod fuel for equivalent LER. The maximum fuel temperature increase as a function of LER increase is also significantly less for ICAF fuel. As a result, the sheath stress induced by an equivalent power increase is approximately six times less for ICAF fuel than solid-rod fuel. This suggests that the power-increase thresholds to failure (due to stress-corrosion cracking) for ICAF fuel should be well above those for solid-rod fuel, providing improvement in operation flexibility and safety.

  6. Performance comparison of metallic, actinide burning fuel in lead-bismuth and sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Weaver, K.D.; Herring, J.S.; Macdonald, P.E.

    2001-01-01

    Various methods have been proposed to ''incinerate'' or ''transmute'' the current inventory of transuranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years. (author)

  7. Safety analysis on tokamak helium cooling slab fuel fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Wei Renjie; Jian Hongbing

    1992-01-01

    The thermal analyses for steady state, depressurization and total loss of flow in the tokamak helium cooling slab fuel element fusion-fission hybrid reactor are presented. The design parameters, computed results of HYBRID program and safety evaluation for conception design are given. After all, it gives some recommendations for developing the design

  8. CFD modelling of cooling channel geometry of PEM fuel cell for ...

    African Journals Online (AJOL)

    In this study, a numerical investigation was carried out to deter mine the impact of cooling channel geometry in combination with temperature dependent operating parameters on thermal management and overall performance of a PEM fuel cell system. The evaluation is performed using a computational fluid dynamics ...

  9. Control and experimental characterization of a methanol reformer for a 350W high temperature polymer electrolyte membrane fuel cell system

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Kær, Søren Knudsen; Jensen, Hans-Christian Becker

    suited for reformer systems, where high CO tolerance is required. This enables the use fuels based on e.g. liquid alcohols. This work presents the control strategies of a methanol refoermer for a 350W HTPEM FC system. The system examined is the Serenergy H3-350 Mobile Battery Charger, an integrated......High temperature polymer electrolyte membrane(HTPEM) fuel cells offer many advantages due to their increased operating tempera-tures compared to similar Nafion-based membrane tech-nologies, that rely on the conductive abilities of liquid water. The polybenzimidazole (PBI) membranes are especially...

  10. Experimental evaluation of a Pt based heat exchanger methanol reformer for a HTPEM fuel cell

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Nielsen, Mads Pagh; Kær, Søren Knudsen

    2007-01-01

    The storage of hydrogen in hydrogen consuming applications is often inconvenient because of the very low density of hydrogen even at high pressures (0.014 kg/L @ 300 bar) or cryogenically (0.043 kg/L). Much higher volumetric energy densities can be achieved using liquid hydrocarbons as e.g. metha...... (up to 1-2%). This work examines the possibility of using a catalyst coated plate heat exchanger for the reforming process of methanol....

  11. Block fuel element for gas-cooled high temperature reactors

    International Nuclear Information System (INIS)

    Hrovat, M.F.

    1978-01-01

    The invention concerns a block fuel element consisting of only one carbon matrix which is almost isotropic of high crystallinity into which the coated particles are incorporated by a pressing process. This block element is produced under isostatic pressure from graphite matrix powder and coated particles in a rubber die and is subsequently subjected to heat treatment. The main component of the graphite matrix powder consists of natural graphite powder to which artificial graphite powder and a small amount of a phenol resin binding agent are added

  12. Methods for studying fuel management in advanced gas cooled reactors

    International Nuclear Information System (INIS)

    Buckler, A.N.; Griggs, C.F.; Tyror, J.G.

    1971-07-01

    The methods used for studying fuel and absorber management problems in AGRs are described. The basis of the method is the use of ARGOSY lattice data in reactor calculations performed at successive time steps. These reactor calculations may be quite crude but for advanced design calculations a detailed channel-by-channel representation of the whole core is required. The main emphasis of the paper is in describing such an advanced approach - the ODYSSEUS-6 code. This code evaluates reactor power distributions as a function of time and uses the information to select refuelling moves and determine controller positions. (author)

  13. Mechanical design issues and resolutions of a dual cooled fuel for the OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung-Kyu, E-mail: hkkim1@kaeri.re.kr [Innovative Nuclear Fuel Division, Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong, Daejeon, 305-353 (Korea, Republic of); Kim, Jae-Yong; Yoon, Kyung-Ho [Innovative Nuclear Fuel Division, Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong, Daejeon, 305-353 (Korea, Republic of)

    2011-06-15

    Highlights: > Thickness of outer cladding tube is determined by using the elastic buckling criterion. > Growth difference of the inner and outer claddings will not cause fuel rod bowing. > Structural components are designed without a drastic change of the conventional ones. - Abstract: A dual cooled fuel is recently brought into focus due to its potential of considerable power uprating. The purpose of present work is to realize the innovative concept of a dual cooled fuel to be a fuel assembly structure compatible with the OPR-1000 system. Under the framework, the critical issues such as the outer cladding thickness and the growth difference of the inner and outer cladding tubes are dealt with in this paper. We designed the thickness of outer cladding tube by using the elastic buckling criterion and safety factor analysis. From the concern of the inner cladding's bowing during irradiation, it was suggested that the outer cladding would grow more than the inner one by applying different microstructures to the inner and outer cladding tubes. It was noted that the gap between fuel rods would not be narrowed further during the different irradiation growth. The structural components such as fuel rod supporting structure, top and bottom end pieces and guide tubes could be designed without a drastic change of those of the conventional fuel. Candidate designs of the components are also presented.

  14. Stress Linearization and Strength Evaluation of the BEP's Flow Plates for a Dual Cooled Fuel Assembly

    International Nuclear Information System (INIS)

    Kim, Jae Yong; Yoon, Kyung Ho; Kang, Heung Seok; Lee, Young Ho; Lee, Kang Hee; Kim, Hyung Kyu

    2009-01-01

    A fuel assembly is composed of 5 major components, such as a top end piece (TEP), a bottom end piece (BEP), spacer grids (SGs), guide tubes (GTs) and an instrumentation tube (IT) and fuel rods (FRs). There are no ASME criteria about all components except for a TEP/BEP. The TEP/BEP should satisfy stress intensity limits in case of condition A and B of ASME, Section III, Division 1 . Subsection NB. In a dual cooled fuel assembly, the array and position of fuels are changed from those of a conventional PWR fuel assembly to achieve a power uprating. The flow plates of top/bottom end pieces (TEP/BEP) have to be modified into proper shape to provide flow holes to direct the heated coolant into/out of the fuel assembly but structural intensity of these plates within a 22.241 kN axial loading should satisfy Tresca stress limits in ASME code. In this paper, stress linearization procedure and strength evaluation of a newly designed BEP for the dual cooled fuel assembly are described

  15. Irradiation performance of U-Pu-Zr metal fuels for liquid-metal-cooled reactors

    International Nuclear Information System (INIS)

    Tsai, H.; Cohen, A.B.; Billone, M.C.; Neimark, L.A.

    1994-10-01

    This report discusses a fuel system utilizing metallic U-Pu-Zr alloys which has been developed for advanced liquid metal-cooled reactors (LMRs). Result's from extensive irradiation testing conducted in EBR-II show a design having the following key features can achieve both high reliability and high burnup capability: a cast nominally U-20wt %Pu-10wt %Zr slug with the diameter sized to yield a fuel smear density of ∼75% theoretical density, low-swelling tempered martensitic stainless steel cladding, sodium bond filling the initial fuel/cladding gap, and an as-built plenum/fuel volume ratio of ∼1.5. The robust performance capability of this design stems primarily from the negligible loading on the cladding from either fuel/cladding mechanical interaction or fission-gas pressure during the irradiation. The effects of these individual design parameters, e.g., fuel smear density, zirconium content in fuel, plenum volume, and cladding types, on fuel element performance were investigated in a systematic irradiation experiment in EBR-II. The results show that, at the discharge burnup of ∼11 at. %, variations on zirconium content or plenum volume in the ranges tested have no substantial effects on performance. Fuel smear density, on the other hand, has pronounced but countervailing effects: increased density results in greater cladding strain, but lesser cladding wastage from fuel/cladding chemical interaction

  16. Improving Fuel Cycle Design and Safety Characteristics of a Gas Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Rooijen, W.F.G. van

    2006-01-01

    The Gas Cooled Fast Reactor (GCFR)is one of the Generation IV reactor concepts. This concept specifically targets sustainability of nuclear power generation. In nuclear reactors fertile material is converted to fissile fuel. If the neutrons inducing fission are highly energetic, the opportunity exists to convert more than one fertile nucleus per fission, thereby effectively breeding new nuclear fuel. Reactors operating on this principle are called ‘Fast Breeder Reactor’. Since natural uranium contains 99.3%of the fertile isotope 238 U, breeding increases the energy harvested from the nuclear fuel. If nuclear energy is to play an important role as a source of energy in the future, fast breeder reactors are essential for breeding nuclear fuel. Fast neutrons are also more efficient to destruct heavy (Minor Actinide, MA) isotopes, such as Np, Am and Cm isotopes, which dominate the long-term radioactivity of nuclear waste. So the waste life-time can be shortened if the MA nuclei are destroyed. An important prerequisite of sustainable nuclear energy is the closed fuel cycle, where only fission products are discharged to a final repository, and all Heavy Metal (HM) are recycled. The reactor should breed just enough fissile material to allow refueling of the same reactor, adding only fertile material to the recycled material. Other key design choices are highly efficient power conversion using a direct cycle gas turbine, and better safety through the use of helium, a chemically inert coolant which cannot have phase changes in the reactor core. Because the envisaged core temperatures and operating conditions are similar to thermal-spectrum High Temperature Reactor (HTR) concepts, the research for this thesis initially focused on a design based on existing HTR fuel technology: coated particle fuel, assembled into fuel assemblies. It was found that such a fuel concept could not meet the Generation IV criteria set for GCFR: self-breeding is difficult, the temperature

  17. Fail-safe system for activity cooled supersonic and hypersonic aircraft. [using liquid hydrogen fuel

    Science.gov (United States)

    Jones, R. A.; Braswell, D. O.; Richie, C. B.

    1975-01-01

    A fail-safe-system concept was studied as an alternative to a redundant active cooling system for supersonic and hypersonic aircraft which use the heat sink of liquid-hydrogen fuel for cooling the aircraft structure. This concept consists of an abort maneuver by the aircraft and a passive thermal protection system (TPS) for the aircraft skin. The abort manuever provides a low-heat-load descent from normal cruise speed to a lower speed at which cooling is unnecessary, and the passive TPS allows the aircraft skin to absorb the abort heat load without exceeding critical skin temperature. On the basis of results obtained, it appears that this fail-safe-system concept warrants further consideration, inasmuch as a fail-safe system could possibly replace a redundant active cooling system with no increase in weight and would offer other potential advantages.

  18. Impact of the water symmetry factor on humidification and cooling strategies for PEM fuel cell stacks

    Energy Technology Data Exchange (ETDEWEB)

    Picot, D; Metkemeijer, R; Bezian, J J; Rouveyre, L [Centre d` Energetique, Ecole des Mines de Paris, 06 - Sophia Antipolis (France)

    1998-10-01

    In this paper, experimental water and thermal balances with three proton exchange membrane fuel cells (PEMFC) are proposed. On the test facility of Ecole des Mines de Paris, three De Nora SPA fuel cell stacks have been successfully studied: An 1 kW{sub e} prototype using Nafion {sup trademark} 117, a 5 and 10 kW{sub e} module using Nafion {sup trademark} 115. The averaged water symmetry factor determines strategies to avoid drying membrane. So, we propose analytical solutions to find compromises between humidification and cooling conditions, which determines outlet temperatures of gases. For transport applications, the space occupied by the power module must be reduced. One of the main efforts consists in decreasing the operative pressure. Thus, if adequate cooling power is applied, we show experimentally and theoretically the possibility to use De Nora PEM fuel cells with low pressure, without specific external humidification. (orig.)

  19. Loss of cooling accident simulation of nuclear power station spent-fuel pool

    Energy Technology Data Exchange (ETDEWEB)

    Lee, M.; Liang, K-S., E-mail: mlee@ess.nthu.edu.tw, E-mail: ksliang_1@hotmail.com [National Tsing Hua Univ., Hsinchu, Taiwan (China); Lin, K-Y., E-mail: syrup760914@gmail.com [Taiwan Power Company, Taiwan (China)

    2014-07-01

    The core melt down accident of Fukushima Nuclear Power Station on March 11th, 2011 alerted nuclear industry that the long term loss of cooling of spent fuel pool may need some attention. The target plant analyzed is the Chinshan Nuclear Power Station of Taiwan Power Company. The 3-Dimensional RELAP5 input deck of the spent fuel pool of the station is built. The results indicate that spent fuel of Chinshan Nuclear Power Station is uncovered at 6.75 days after an accident of loss cooling takes place and cladding temperature rises above 2,200{sup o}F around 8 days. The time is about 13 hours earlier than the results predicted using simple energy balance method. The results also show that the impact of Counter Current Flow Limitation (CCFL) and radiation heat transfer model is marginal. (author)

  20. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Mcwilliams, A. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  1. Crossflow characteristics of flange type fuel element for very high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Takizuka, Takakazu; Kaburaki, Hideo; Suzuki, Kunihiko; Nakamura, Masahide.

    1987-01-01

    Fuel element design incorporating mating flanges at block end faces has the potential to improve thermal hydraulic performance of a VHTR (very high temperature gas-cooled reactor) core. As part of research and development efforts to establish flange type fuel element design, experiments and analyses were carried out on crossflow through interface gap between elements. Air at atmospheric pressure and ambient temperature was used as a fluid. Crossflow loss coefficient factors were obtained with three test models, having different flange mating clearances, for various interface gap configurations, gap widths and block misalignments. It was found that crossflow loss coefficient factors for flange type fuel element were much larger than those for conventional flat-faced element. Numerical analyses were also made using a simple model devised to represent the crossflow path at the fuel element interface. The close agreement between numerical results and experimental data indicated that this model could predict well the crossflow characteristics of the flange type fuel element. (author)

  2. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  3. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

    2011-03-01

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  4. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  5. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    International Nuclear Information System (INIS)

    Armstrong, J.; Hamilton, H.; Hyland, B.

    2013-01-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  6. CFD Analysis of the Fuel Temperature in High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    In, W. K.; Chun, T. H.; Lee, W. J.; Chang, J. H.

    2005-01-01

    High temperature gas-cooled reactors (HTGR) have received a renewed interest as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor (PBR) and a prismatic modular reactor (PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both PBR and PMR. The objective of this study is to predict the fuel temperature distributions in PBR and PMR using a computational fluid dynamics(CFD) code, CFX-5. The reference reactor designs used in this analysis are PBMR400 and GT-MHR600

  7. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

    2014-10-15

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

  8. Modelling of Water Cooled Fuel Including Design Basis and Severe Accidents. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2015-11-01

    The demands on nuclear fuel have recently been increasing, and include transient regimes, higher discharge burnup and longer fuel cycles. This has resulted in an increase of loads on fuel and core internals. In order to satisfy these demands while ensuring compliance with safety criteria, new national and international programmes have been launched and advanced modelling codes are being developed. The Fukushima Daiichi accident has particularly demonstrated the need for adequate analysis of all aspects of fuel performance to prevent a failure and also to predict fuel behaviour were an accident to occur.This publication presents the Proceedings of the Technical Meeting on Modelling of Water Cooled Fuel Including Design Basis and Severe Accidents, which was hosted by the Nuclear Power Institute of China (NPIC) in Chengdu, China, following the recommendation made in 2013 at the IAEA Technical Working Group on Fuel Performance and Technology. This recommendation was in agreement with IAEA mid-term initiatives, linked to the post-Fukushima IAEA Nuclear Safety Action Plan, as well as the forthcoming Coordinated Research Project (CRP) on Fuel Modelling in Accident Conditions. At the technical meeting in Chengdu, major areas and physical phenomena, as well as types of code and experiment to be studied and used in the CRP, were discussed. The technical meeting provided a forum for international experts to review the state of the art of code development for modelling fuel performance of nuclear fuel for water cooled reactors with regard to steady state and transient conditions, and for design basis and early phases of severe accidents, including experimental support for code validation. A round table discussion focused on the needs and perspectives on fuel modelling in accident conditions. This meeting was the ninth in a series of IAEA meetings, which reflects Member States’ continuing interest in nuclear fuel issues. The previous meetings were held in 1980 (jointly with

  9. Fuel Fraction Analysis of 500 MWth Gas Cooled Fast Reactor with Nitride (UN-PuN) Fuel without Refueling

    Science.gov (United States)

    Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

    2017-01-01

    Nuclear Power Plant (NPP) is one of candidates which can support electricity demand in the world. The Generation IV NPP has fourth main objective, i.e. sustainability, economics competitiveness, safety and reliability, and proliferation and physical protection. One of Gen-IV reactor type is Gas Cooled Fast Reactor (GFR). In this study, the analysis of fuel fraction in small GFR with nitride fuel has been done. The calculation was performed by SRAC code, both Pij and CITATION calculation. SRAC2002 system is a code system applicable to analyze the neutronics of variety reactor type. And for the data library used JENDL-3.2. The step of SRAC calculation is fuel pin calculated by Pij calculation until the data homogenized, after it homogenized we calculate core reactor. The variation of fuel fraction is 40% up to 65%. The optimum design of 500MWth GFR without refueling with 10 years burn up time reach when radius F1:F2:F3 = 50cm:30cm:30cm and height F1:F2:F3 = 50cm:40cm:30cm, variation percentage Plutonium in F1:F2:F3 = 7%:10%:13%. The optimum fuel fraction is 41% with addition 2% Plutonium weapon grade mix in the fuel. The excess reactivity value in this case 1.848% and the k-eff value is 1.01883. The high burn up reached when the fuel fraction is low. In this study 41% fuel fraction produce faster fissile fuel, so it has highest burn-up level than the other fuel fraction.

  10. Design of the Flow Plates for a Dual Cooled Fuel Assembly

    International Nuclear Information System (INIS)

    Kim, Jae Yong; Yoon, Kyung Ho; Lee, Young Ho; Lee, Kang Hee; Kim, Hyung Kyu

    2009-01-01

    In a dual cooled fuel assembly, the array and position of fuels are changed from those of a conventional PWR fuel assembly to achieve a power uprating. The flow plate provides flow holes to direct the heated coolant into/out of the fuel assembly and structural intensity to insure that the fuel rod is axially restrained within the spacer grids. So, flow plates of top/bottom end pieces (TEP/BEP) have to be modified into proper shape. Because the flow holes' area of a flow plate affects pressure drop, the flow holes' area must be larger than/equal to that of conventional flow plates. And design criterion of the TEP/BEP says that the flow plate should withstand a 22.241 kN axial load during handling lest a calculated stress intensity should exceed the Condition I allowable stress. In this paper, newly designed flow plates of a TEP/BEP are suggested and stress analysis is conducted to evaluate strength robustness of the flow plates for the dual cooled fuel assembly

  11. KUEBEL. A Fortran program for computation of cooling-agent-distribution within reactor fuel-elements

    International Nuclear Information System (INIS)

    Inhoven, H.

    1984-12-01

    KUEBEL is a Fortran-program for computation of cooling-agent-distribution within reactor fuel-elements or -zones of theirs. They may be assembled of max. 40 cooling-channels with laminar up to turbulent type of flow (respecting Reynolds' coefficients up to 2.0E+06) at equal pressure loss. Flow-velocity, dynamic flow-, contraction- and friction-losses will be calculated for each channel and for the total zone. Other computations will present mean heat-up of cooling-agent, mean outlet-temperature of the core, boiling-temperature and absolute pressure at flow-outlet. All characteristic coolant-values, including the factor of safety for flow-instability of the most-loaded cooling gap are computed by 'KUEBEL' too. Absolute pressure at flow-outlet or is-factor may be defined as dependent or independent variables of the program alternatively. In latter case 3 variations of solution will be available: Adapted flow of cooling-agent, inlet-temperature of the core and thermal power. All calculations can be done alternatively with variation of parameters: flow of cooling-agent, inlet-temperature of the core and thermal power, which are managed by the program itself. 'KUEBEL' is able to distinguish light- and heavy-water coolant, flow-direction of coolant and fuel elements with parallel, rectangular, respectively concentric, cylindrical shape of their gaps. Required material specifics are generated by the program. Segments of fuel elements or constructively unconnected gaps can also be computed by means of interposition of S.C. 'phantom channels'. (orig.) [de

  12. Challenges and innovative technologies on fuel handling systems for future sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Chassignet, Mathieu; Dumas, Sebastien; Penigot, Christophe; Prele, Gerard; Capitaine, Alain; Rodriguez, Gilles; Sanseigne, Emmanuel; Beauchamp, Francois

    2011-01-01

    The reactor refuelling system provides the means of transporting, storing, and handling reactor core subassemblies. The system consists of the facilities and equipment needed to accomplish the scheduled refuelling operations. The choice of a FHS impacts directly on the general design of the reactor vessel (primary vessel, storage, and final cooling before going to reprocessing), its construction cost, and its availability factor. Fuel handling design must take into account various items and in particular operating strategies such as core design and management and core configuration. Moreover, the FHS will have to cope with safety assessments: a permanent cooling strategy to prevent fuel clad rupture, plus provisions to handle short-cooled fuel and criteria to ensure safety during handling. In addition, the handling and elimination of residual sodium must be investigated; it implies specific cleaning treatment to prevent chemical risks such as corrosion or excess hydrogen production. The objective of this study is to identify the challenges of a SFR fuel handling system. It will then present the range of technical options incorporating innovative technologies under development to answer the GENERATION IV SFR requirements. (author)

  13. Tools for designing the cooling system of a proton exchange membrane fuel cell

    International Nuclear Information System (INIS)

    Soupremanien, Ulrich; Le Person, Stéphane; Favre-Marinet, Michel; Bultel, Yann

    2012-01-01

    Proton exchange membrane fuel cell (PEMFC) requires a careful management of the heat distribution inside the stack. The proton exchange membrane is the most sensitive element of this thermal management and it must operate under specific conditions in order to increase the lifetime and also the output power of the fuel cell. These last decades, the enhancement of the output power of the PEMFC has led the manufacturers to greatly improve the heat transfer effectiveness for cooling such systems. In addition, homogenizing the bipolar plate temperature increases the lifetime of the system by limiting the occurrence of strong thermal gradients. In this context, using a fluid in boiling conditions to cool down the PEMFC seems to be very suitable for this purpose. In order to compare the thermal performances between a coolant used in single-phase flow or in boiling flow conditions, we have built an experimental set-up allowing the investigation of cooling flows for these two conditions. Moreover, the geometry of the cooling channels is one of the key parameters which allows the improvement of the thermal performances. Indeed, the size or the aspect ratio of these channels could be designed in order to decrease the thermal system response. The sizing of the fuel cell cooling system is of paramount importance in boiling flow conditions because it can modify, not only the pressure losses along the channel and the heat transfer coefficient like in a single-phase flow but also, the onset of nucleate boiling (ONB) and the dryout point or critical heat flux (CHF). Thus, in order to understand some heat transfer mechanisms, which are geometry-dependent, a parametric study was completed by considering flows in four different rectangular channels. Finally, this study allows a better insight on the optimization of the geometrical parameters which improve the thermal performances of a PEMFC, from a cooling strategy aspect point of view. - Highlights: ► Parameters for the using of a

  14. Basic concept of fuel safety design and assessment for sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Nakae, Nobuo; Baba, Toshikazu; Kamimura, Katsuichiro

    2013-03-01

    'Philosophy in Safety Evaluation of Fast Breeder Reactors' was published as a guideline for safety design and safety evaluation of Sodium-Cooled Fast Reactor in Japan. This guideline points out that cladding creep and swelling due to internal pressure should be taken into account since the fuel is used under high temperature and high burnup, and that fuel assembly deformation and the prevention from coolant channel blockage should be taken into account in viewpoints of nuclear and thermal hydraulic design. However, the requirements including their criteria and evaluation items are not described. Two other domestic guidelines related to core design are applied for fuel design of fast reactor, but the description is considered to not be enough to practically use. In addition, technical standard for nuclear fuel used in power reactors is also applied for fuel inspection. Therefore, the technical standard and guideline for fuel design and safety evaluation are considered to be very important issue for nuclear safety regulation. This document has been developed according to the following steps: The guidelines and the technical standards, which are prepared in foreign countries and international organization, were reviewed. The technical background concerning fuel design and safety evaluation for fast reactor was collected and summarized in the world wide scale. The basic concept of fuel safety design and assessment for sodium-cooled fast reactor was developed by considering a wide range of views of the specialists in Japan. In order to discuss the content with foreign specialists IAEA Consultancy Meetings have been held on January, 2011 and January, 2012. The participants of the meeting came from USA, UK, EC, India, China and South Korea. The specialists of IAEA and JNES were also joined. Although this document is prepared for application to 'Monju'(prototype LMFR), it may be applied to experimental, demonstration and commercial types of LMFR after revising it by taking

  15. AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Gorgemans, J.; Mulhollem, L.; Glavin, J.; Pfister, A.; Conway, L.; Schulz, T.; Oriani, L.; Cummins, E.; Winters, J. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first level of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all

  16. Performance of water cooled nuclear power reactor fuels in India – Defects, failures and their mitigation

    International Nuclear Information System (INIS)

    Ganguly, Chaitanyamoy

    2015-01-01

    Water cooled and moderated nuclear power reactors account for more than 95% of the operating reactors in the world today. Light water reactors (LWRs) consisting of pressurized water reactor (PWR), their Russian counterpart namely VVER and boiling water reactor (BWR) will continue to dominate the nuclear power market. Pressurized heavy water reactor (PHWR), also known as CANDU, is the backbone of the nuclear power program in India. Updates on LWR and PHWR fuel performance are being periodically published by IAEA, OECD-NEA and the World Nuclear Association (WNA), highlighting fuel failure rate and the mitigation of fuel defects and failures. These reports clearly indicate that there has been significant improvement in in – pile fuel performance over the years and the present focus is to achieve zero fuel failure in high burn up and high performance fuels. The present paper summarizes the status of PHWR and LWR fuel performance in India, highlighting the manufacturing and the related quality control and inspection steps that are being followed at the PHWR fuel fabrication plant in order to achieve zero manufacturing defect which could contribute to achieving zero in – pile failure rate in operating and upcoming PHWR units in India. (author)

  17. The continuous fuel cycle model and the gas cooled fast reactor

    International Nuclear Information System (INIS)

    Christie, Stuart; Lathouwers, Danny; Kloosterman, Jan Leen; Hagen, Tim van der

    2011-01-01

    The gas cooled fast reactor (GFR) is one of the generation IV designs currently being evaluated for future use. It is intended to behave as an isobreeder, producing the same amount of fuel as it consumes during operation. The actinides in the fuel will be recycled repeatedly in order to minimise the waste output to fission products only. Striking the balance of the fissioning of various actinides against transmutation and decay to achieve these goals is a complex problem. This is compounded by the time required for burn-up modelling, which can be considerable for a single cycle, and even longer for studies of fuel evolution over many cycles. The continuous fuel cycle model approximates the discrete steps of loading, operating and unloading a reactor as continuous processes. This simplifies the calculations involved in simulating the behaviour of the fuel, reducing the time needed to model the changes to the fuel composition over many cycles. This method is used to study the behaviour of GFR fuel over many cycles and compared to results obtained from direct calculations. The effects of varying fuel cycle properties such as feed material, recycling of additional actinides and reprocessing losses are also investigated. (author)

  18. On-Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactors

    International Nuclear Information System (INIS)

    Hawari, Ayman I.; Bourham, Mohamed A.

    2010-01-01

    Very High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (∼ 1-mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4%-10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.

  19. Standard casks for the transport of LWR spent fuel. Storage/transport casks for long cooled spent fuel

    International Nuclear Information System (INIS)

    Blum, P.; Sert, G.; Gagnon, R.

    1983-01-01

    During the past decade, TRANSNUCLEAIRE has developed, licensed and marketed a family of standard casks for the transport of spent fuel from LWR reactors to reprocessing plants and the ancillary equipments necessary for their operation and transport. A large number of these casks are presently used for European and intercontinental transports and manufactured under TRANSNUCLEAIRE supervision in different countries. The main advantages of these casks are: - large payload for considered modes of transport, - moderate cost, - reliability due to the large experience gained by TRANSNUCLEAIRE as concerns fabrication and operation problems, - standardization faciliting fabrication, operation and spare part supply. Recently, TRANSNUCLEAIRE also developed a new generation of casks for the dry storage and occasional transport of LWR spent fuel which has been cooled for 5 years or 7 years in case of consolidated fuel rods. These casks have an optimum payload which takes into account the shielding requirements and the weight limitations at most sites. This paper deals more particularly with the TN 24 model which exists in 4 versions among which one for 24 PWR 900 fuel assemblies and another one for the consolidated fuel rods from 48 of same fuel assemblies

  20. Nuclear design for high temperature gas cooled reactor (GTHTR300C) using MOX fuel

    International Nuclear Information System (INIS)

    Mouri, Tomoaki; Kunitomi, Kazuhiko

    2008-01-01

    A design study of the hydrogen cogeneration high temperature gas cooled reactor (GTHTR300C) that can produce both electricity and hydrogen has been carried out in Japan Atomic Energy Agency. The GTHTR300C is the system with thermal power of 600MW and reactor outlet temperature of 950degC, which is expected to supply the hydrogen to fuel cell vehicles after 2020s. In future, the full deployment of fast reactor cycle without natural uranium will demand the use of Mixed-Oxide (MOX) fuels in the GTHTR300C. Therefore, a nuclear design was performed to confirm the feasibility of the reactor core using MOX fuels. The designed reactor core has high performance and meets safety requirements. In this paper, the outline of the GTHTR300C and the nuclear design of the reactor core using MOX fuels are described. (author)

  1. Optimization of the fuel assembly for the Canadian SuperCritical Water-cooled Reactor (SCWR)

    Energy Technology Data Exchange (ETDEWEB)

    French, C., E-mail: Corey.French@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada); Bonin, H.; Chan, P.K. [Royal Military College of Ontario, Kingston, Ontario (Canada)

    2013-07-01

    An approach to develop a parametric optimization tool to support the Canadian Supercritical Water-cooled Reactor (SCWR) fuel design is presented in this work. The 2D benchmark lattices for 78-pin and 64-pin fuel assemblies are used as the initial models from which fuel performance and subsequent optimization stem from. A tandem optimization procedure is integrated which employs the steepest descent method. The physics codes WIMS-AECL, MCNP6 and SERPENT are used to calculate and verify select performance factors. The results are used as inputs to an optimization algorithm that yield optimal fresh fuel isotopic composition and lattice geometry. Preliminary results on verifications of infinite lattice reactivity are demonstrated in this paper. (author)

  2. A design study of high breeding ratio sodium cooled metal fuel core without blanket fuels

    International Nuclear Information System (INIS)

    Kobayashi, Noboru; Ogawa, Takashi; Ohki, Shigeo; Mizuno, Tomoyasu; Ogata, Takanari

    2009-01-01

    The metal fuel core is superior to the mixed oxide fuel core because of its high breeding ratio and compact core size resulting from hard neutron spectrum and high heavy metal densities. Utilizing these characteristics, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void worth of less than 8 $, a core height of less than 150 cm, the maximum cladding temperature of 650degC, and the maximum fuel pin bundle pressure drop of 0.4 MPa. The breeding ratio of the resultant core was 1.34 with 6wt% zirconium content fuel. Applying 3wt% zirconium content fuel enhanced the breeding ratio up to 1.40. (author)

  3. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    Science.gov (United States)

    Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

    1993-01-01

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

  4. Response of unirradiated and irradiated PWR fuel rods tested under power-cooling-mismatch conditions

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Quapp, W.J.; Martinson, Z.R.; McCardell, R.K.; Mehner, A.S.

    1978-01-01

    This report summarizes the results from the single-rod power-cooling-mismatch (PCM) and irradiation effects (IE) tests conducted to date in the Power Burst Facility (PBF) at the U.S. DOE Idaho National Engineering Laboratory. This work was performed for the U.S. NRC under contact to the Department of Energy. These tests are part of the NRC Fuel Behavior Program, which is designed to provide data for the development and verification of analytical fuel behavior models that are used to predict fuel response to abnormal or postulated accident conditions in commercial LWRs. The mechanical, chemical and thermal response of both previously unirradiated and previously irradiated LWR-type fuel rods tested under power-cooling-mismatch condition is discussed. A brief description of the test designs is presented. The results of the PCM thermal-hydraulic studies are summarized. Primary emphasis is placed on the behavior of the fuel and cladding during and after stable film boiling. (orig.) [de

  5. The encapsulation of Magnox type fuel elements for extended storage in cooling ponds

    International Nuclear Information System (INIS)

    Baker, D.W.C.; Burt, G.A.

    1978-01-01

    A method of encapsulating spent fuel elements in a protective plastics medium to enable them to be stored for protracted periods under water, without risk of further significant corrosion, has been developed. It is visualised that the elements after discharge from the reactor would be allowed to cool under water for a period of at least 100 days and would then be encapsulated while remaining immersed. A suitable two pack system based on a solvent free epoxy resin cured with an aromatic amine adduct has been identified. The equipment and processes which have been developed for handling, conditioning and encapsulating the fuel are described. (author)

  6. Investigation of TIG welding characteristics with a dual cooled rod for the fuel irradiation test

    International Nuclear Information System (INIS)

    Kim, Soo Sung; Kim, Hyung Kyu

    2008-01-01

    To establish the fabrication process, and for satisfying the requirements of the irradiation test, an TIG(Tungsten Inert Gas) welding machine for the dual cooled rods specimens was developed, and the preliminary welding experiments were performed to optimize the welding process conditions. Cladding tubes of 15.9 and 9 mm for the outer and inner diameters, respectively with a 0.57 mm thickness and end caps were used for the specimens. This paper describes the experimental results of the TIG welds and the micrograph examinations of the TIG welded specimens corresponding to various welding conditions for the dual cooled fuel irradiation test. The investigations revealed that the present TIG process satisfied the requirements for the fuel irradiation test in the HANARO research reactor

  7. Burn-up measurements of LEU fuel for short cooling times

    International Nuclear Information System (INIS)

    Pereda B, C.; Henriquez A, C.; Klein D, J.; Medel R, J.

    2005-01-01

    The measurements presented in this work were made essentially at in-pool gamma-spectrometric facility, installed inside of the secondary pool of the RECH-1 research reactor, where the measured fuel elements are under 2 meters of water. The main reason for using the in-pool facility was because of its capability to measure the burning of fuel elements without having to wait so long, that is with only 5 cooling days, which are the usual times between reactor operations. Regarding these short cooling times, this work confirms again the possibility of using the 95 Zr as a promising burnup monitor, in spite of the rough approximations used to do it. These results are statistically reasonable within the range calculated using codes. The work corroborates previous results, presented in Santiago de Chile, and it suggests future improvements in that way. (author)

  8. Investigations on the thermal-hydraulics of a natural circulation cooled BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kok, H.V.; Hagen, T.H.J.J. van der; Mudde, R.F. [Delft Univ. of Technology (Netherlands)

    1995-09-01

    A scaled natural circulation loop facility has been built after the Dodewaard Boiling Water Reactor, which is the only operating natural circulation cooled BWR in the world. The loop comprises one fuel assembly, a riser with a downcomer and a condenser with a cooling system. Freon-12 is used as a scaling liquid. This paper reports on the first measurements done with this facility. Quantities like the circulation flow, carry-under and the void-fraction have been measured as a function of power, pressure, liquid level, riser length, condensate temperature and friction factors. The behavior of the circulation flow can be understood by considering the driving force. Special attention has been paid to the carry-under, which has been shown to have a very important impact on the dynamics of a natural circulation cooled BWR.

  9. Processing of FRG high-temperature gas-cooled reactor fuel elements at General Atomic under the US/FRG cooperative agreement for spent fuel elements

    International Nuclear Information System (INIS)

    Holder, N.D.; Strand, J.B.; Schwarz, F.A.; Drake, R.N.

    1981-11-01

    The Federal Republic of Germany (FRG) and the United States (US) are cooperating on certain aspects of gas-cooled reactor technology under an umbrella agreement. Under the spent fuel treatment development section of the agreement, both FRG mixed uranium/ thorium and low-enriched uranium fuel spheres have been processed in the Department of Energy-sponsored cold pilot plant for high-temperature gas-cooled reactor (HTGR) fuel processing at General Atomic Company in San Diego, California. The FRG fuel spheres were crushed and burned to recover coated fuel particles suitable for further treatment for uranium recovery. Successful completion of the tests described in this paper demonstrated certain modifications to the US HTGR fuel burining process necessary for FRG fuel treatment. Results of the tests will be used in the design of a US/FRG joint prototype headend facility for HTGR fuel

  10. Fuel cycles and advanced core designs for the Gas-Cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Simon, R.H.; Hamilton, C.J.; Hunter, R.S.

    1982-01-01

    Studies indicate that a 1200 MW(e) Gas-Cooled Fast Breeder Reactor could achieve compound system doubling times of under ten years when using advanced oxide or carbide fuels. In addition, when thorium is used in the breeding blankets, enough U-233 can be generated in each GCFR to supply several advanced converter reactors with fissionable material and this symbiotic relationship could provide energy for the world for centuries. (author)

  11. Optimization of advanced gas-cooled reactor fuel performance by a stochastic method

    International Nuclear Information System (INIS)

    Parks, G.T.

    1987-01-01

    A brief description is presented of a model representing the in-core behaviour of a single advanced gas-cooled reactor fuel channel, developed specifically for optimization studies. The performances of the only suitable Numerical Algorithms Group (NAG) library package and a Metropolis algorithm routine on this problem are discussed and contrasted. It is concluded that, for the problem in question, the stochastic Metropolis algorithm has distinct advantages over the deterministic NAG routine. (author)

  12. Technical development and economic valuation of new cooling methods for planar solid oxide fuel cells (SOFC)

    International Nuclear Information System (INIS)

    Thom, F.

    2002-02-01

    A great potential exists for the use of the solid oxide fuel cell technology based on the planar cell design concept. Besides its application as power provider there is a need to supply process heat in the temperature range of 200 to 1200 C for commercial and industrial decentralized facilities. The present study is concerned with the technical development and economic valuation of plant concepts of new fuel cell cooling methods. They can be considered as an alternative to the normal convective cell cooling with air. Besides experimental studies on the natural gas reforming with the SOFC special attention is paid to the process analysis of the power plant carried out with the simulating program PROII. The 200 kWe SOFC is linked with peripheral components such as prereformer, heat exchangers, compressors etc. Developed program subroutine serve to calculate the electrical power output of the fuel cell, the investment costs and the costs of electricity. The study shows clearly that a radiative cell cooling device on basis of an external arranged vaporizer has economic benefits in comparison with the normal air cooling. In this case the possibility is given to run the fuel cell with completely prereformed natural gas. When the internal methane reforming is carried out in excess of the electrochemical demand for hydrogen and carbon monoxide respectively a further cost reduction potential is given. The produced synthesis gas can be used in alternative to the production of power in a gas turbine to supply process steam in the temperature range of 200 to 1200 C. Sensitivity analyses show that a successive use of optimization potentials (e.g. anode structure and operating parameters of the SOFC) leads to a further reduction of the costs of electricity. In the best case the achieved costs of 12 to 13 Pf/kWh are in a range achieved by CHP plants based on engines. (orig.) [de

  13. Numerical simulation of the hydrodynamic behavior of fuel rod with longitudinal cooling fins

    International Nuclear Information System (INIS)

    Naot, D.; Emrani, S.

    1982-01-01

    Four processes which considerably affect the distribution of the local shear stress in turbulent cooling flow along a fuel rod with longitudinal fins are discussed. The effect of boundary layers' development, geometry driven secondary currents, roughness induced lateral motion and geometry imperfections were studied and compared. Turbulence was modeled by an energy-dissipation model with an algebraic stress model. The three-dimensional flow was numerically simulated using a parabolic pressure correction algorithm. (orig.)

  14. EBR-II argon cooling system restricted fuel handling I and C upgrade

    International Nuclear Information System (INIS)

    Start, S.E.; Carlson, R.B.; Gehrman, R.L.

    1995-01-01

    The instrumentation and control of the Argon Cooling System (ACS) restricted fuel handling control system at Experimental Breeder Reactor II (EBR-II) is being upgraded from a system comprised of many discrete components and controllers to a computerized system with a graphical user interface (GUI). This paper describes the aspects of the upgrade including reasons for the upgrade, the old control system, upgrade goals, design decisions, philosophies and rationale, and the new control system hardware and software

  15. Heat transfer analysis in internally-cooled fuel elements by means of a conformal mapping approach

    International Nuclear Information System (INIS)

    Sarmiento, G.S.; Laura, P.A.A.

    1981-01-01

    The present paper deals with an approximate solution of the steady-state heat conduction problem in internally cooled fuel elements of fast breeder reactors. Explicit expressions for the dimensionless temperature distribution in terms of the governing physical and geometrical parameters are determined by means of a coupled conformal mapping-variational approach. The results obtained are found to be in very good agreement with those calculated by means of a finite element code. (orig.)

  16. Characterization of effluents from a high-temperature gas-cooled reactor fuel refabrication plant

    International Nuclear Information System (INIS)

    Judd, M.S.; Bradley, R.A.; Olsen, A.R.

    1975-12-01

    The types and quantities of chemical and radioactive effluents that would be released from a reference fuel refabrication facility for the High-Temperature Gas-Cooled Reactor (HTGR) have been determined. This information will be used to predict the impact of such a facility on the environment, to identify areas where additional development work needs to be done to further identify and quantify effluent streams, and to limit effluent release to the environment

  17. Reducing the fuel temperature for pressure-tube supercritical-water-cooled reactors and the effect of fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: eleodor.nichita@uoit.ca; Kovaltchouk, V., E-mail: vitali.kovaltchouk@uoit.ca

    2015-12-15

    Highlights: • Typical PT-SCWR fuel uses single-region pins consisting of a homogeneous mixture of ThO{sub 2} and PuO{sub 2}. • Using two regions (central for the ThO{sub 2} and peripheral for the PuO{sub 2}) reduces the fuel temperature. • Single-region-pin melting-to-average power ratio is 2.5 at 0.0 MW d/kg and 2.3 at 40 MW d/kg. • Two-region-pin melting-to-average power ratio is 36 at 0.0 MW d/kg and 10.5 at 40 MW d/kg. • Two-region-pin performance drops with burnup due to fissile-element buildup in the ThO{sub 2} region. - Abstract: The Pressure-Tube Supercritical-Water-Cooled Reactor (PT-SCWR) is one of the concepts under investigation by the Generation IV International Forum for its promise to deliver higher thermal efficiency than nuclear reactors currently in operation. The high coolant temperature (>625 K) and high linear power density employed by the PT-SCWR cause the fuel temperature to be fairly high, leading to a reduced margin to fuel melting, thus increasing the risk of actual melting during accident scenarios. It is therefore desirable to come up with a fuel design that lowers the fuel temperature while preserving the high linear power ratio and high coolant temperature. One possible solution is to separate the fertile (ThO{sub 2}) and fissile (PuO{sub 2}) fuel materials into different radial regions in each fuel pin. Previously-reported work found that by locating the fertile material at the centre and the fissile material at the periphery of the fuel pin, the fuel centreline temperature can be reduced by ∼650 K for fresh fuel compared to the case of a homogeneous (Th–Pu)O{sub 2} mixture for the same coolant temperature and linear power density. This work provides a justification for the observed reduction in fuel centreline temperature and suggests a systematic approach to lower the fuel temperature. It also extends the analysis to the dependence of the radial temperature profile on fuel burnup. The radial temperature profile is

  18. Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

    International Nuclear Information System (INIS)

    Mella, R.; Wenman, M.R.

    2013-01-01

    Thermo-mechanical contributions to pellet–clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS’s well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used

  19. Fabrication of uranium alloy fuel slug for sodium-cooled fast reactor by injection casting

    International Nuclear Information System (INIS)

    Jong Hwan Kim; Hoon Song; Ki Hwan Kim; Chan Bock Lee

    2014-01-01

    Metal fuel slugs of U-Zr alloys for a sodium-cooled fast reactor (SFR) have been fabricated using an injection casting method. However, casting alloys containing volatile radioactive constituents such as Am can cause problems in a conventional injection casting method. Therefore, in this study, several injection-casting methods were applied to evaluate the volatility of the metal-fuel elements and control the transport of volatile elements. Mn was selected as a volatile surrogate alloy since it possesses a total vapor pressure equivalent to that of minor actinide-bearing fuels for SFRs. U-10 wt% Zr and U-10 wt% Zr-5 wt% Mn metal fuels were prepared, and the casting processes were evaluated. The casting soundness of the fuel slugs was characterized by gamma-ray radiography and immersion density measurements. Inductively coupled plasma atomic emission spectroscopy was used to determine the chemical composition of fuel slugs. Fuel losses after casting were also evaluated according to the casting conditions. (author)

  20. Quality control of coated fuel particles for high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Kaneko, Mitsunobu

    1987-01-01

    The quality control of the coated fuel particles for high temperature gas-cooled reactors is characterized by the fact that the size of the target product to be controlled is very small, and the quantity is very large. Accordingly, the sampling plan and the method of evaluating the population through satisfically treating the measured data of the samples are the important subjects to see and evaluate the quality of a batch or a lot. This paper shows the fabrication process and the quality control procedure for the coated fuel particles. The development work of a HTGR was started by Japan Atomic Energy Research Institute in 1969, and as for the production technology for coated fuel particles, Nuclear Fuel Industries, Ltd. has continued the development work. The pilot plan with the capacity of about 40 kg/year was established in 1972. The fuel product fabricated in this plant was put to the irradiation experiment and out-of-pile evaluation test. In 1983, the production capacity was expanded to 200 kg/year, and the fuel compacts for the VHTRC in JAERI were produced for two years. The basic fuel design, the fabrication process, the quality control, the process control and the quality assurance are reported. For the commercial product, the studies from the viewpoint of production and quality control costs are required. (Kako, I.)

  1. Feasibility Study on Dual-Cooled Annular Fuel for OPR-1000 Power Uprate

    International Nuclear Information System (INIS)

    Chun, Tae Hyun; In, Wang Kee; Oh, Dong Suck

    2010-04-01

    A dual-cooled annular fuel (DCAF) is a highly promising concept as a high power density fuel for PWR power-uprate. The purpose of this study is to assess a feasibility of 120% core power for OPR-1000 with the DCAF. So the feasibility study were done through the code establishments for annular fuel analysis, evaluations of core physics, thermal-hydraulics and safety analyses at a 120% power with OPR-1000 and the preliminary economic benefits of 20% power-uprate. As results of the analyses, DCAF at 120% power showed sufficient margins available on DNB, PCT and fuel pellet temperature relative to the solid fuel at 100% power. However, judging from an anticipated wide range of the gap conductance variation in inner and outer clearances as fuel burn-up in the reactor core, irradiation behavior of DCAF has to be observed through research reactor test. On the other hand, the nuclear physics parameters like moderator temperature coefficient, power coefficient and so on comply with the technical specifications. An impact of 20% power-uprate on NSSS and BOP was also investigated, and accordingly some components and parts need to be changed were identified. Moreover, the economical benefits from the power-uprate was roughly estimated. It turned out that the power-uprating with DCAF could give an enormous profit even considering the expenses of components and parts to be replaced, additional fuel cycle cost and extended overhaul period

  2. Using Small Capacity Fuel Cells Onboard Drones for Battery Cooling: An Experimental Study

    Directory of Open Access Journals (Sweden)

    Shayok Mukhopadhyay

    2018-06-01

    Full Text Available Recently, quadrotor-based drones have attracted a lot of attention because of their versatility, which makes them an ideal medium for a variety of applications, e.g., personal photography, surveillance, and the delivery of lightweight packages. The flight duration of a drone is limited by its battery capacity. Increasing the payload capacity of a drone requires more current to be supplied by the battery onboard a drone. Elevated currents through a Li-ion battery can increase the battery temperature, thus posing a significant risk of fire or explosion. Li-ion batteries are suited for drone applications, due to their high energy density. There have been attempts to use hydrogen fuel cells onboard drones. Fuel cell stacks and fuel tank assemblies can have a high energy to weight ratio. So, they may be able to power long duration drone flights, but such fuel cell stacks and associated systems, are usually extremely expensive. Hence, this work proposes the novel use of a less expensive, low capacity, metal hydride fuel stick-powered fuel cell stack as an auxiliary power supply onboard a drone. A primary advantage of this is that the fuel sticks can be used to cool the batteries, and a side effect is that this slightly reduces the burden on the onboard Li-ion battery and provides a small increment in flight time. This work presents the results of an experimental study which shows the primary effect (i.e., decrease in battery temperature and the secondary side effect (i.e., a small increment in flight time obtained by using a fuel cell stack. In this work, a metal hydride fuel stick powered hydrogen fuel cell is used along with a Li-ion battery onboard a drone.

  3. The fuel-cladding interfacial friction coefficient in water-cooled reactor fuel rods

    International Nuclear Information System (INIS)

    Smith, E.

    1979-01-01

    A central problem in the development of cladding failure criteria and of effective operational, design or material remedies is to know whether the cladding stress is enhanced significantly near cladding ridges, pellet chips or fuel pellet cracks; the latter may also be coincident with cladding ridges at pellet-pellet interfaces. As regards the fuel pellet crack source of cladding stress concentration, the magnitude of the uranium dioxide-Zircaloy interfacial friction coefficient μ governs the magnitude and distribution of the enhanced cladding stress. Considerable discussion, particularly at a Post-Conference Seminar associated with the SMIRT 4 Conference, has focussed on the value of μ, the author taking the view that it is unlikely to be large (< 0.5). The reasoning behind this view is as follows. A fuel pellet should fracture during a power ramp when the tensile hoop stress within the pellet exceeds the fuel's fracture stress. Since the preferred position for a fuel pellet crack to form is at the fuel-cladding interface midway between existing fuel cracks, where the interfacial shear stress changes sign, the pellet segment size after a power ramp provides a limit to the magnitude of the interfacial shear stresses and consequently to the value of μ. With this argument as a basis, the author's early work used the Gittus fuel rod model, in which there is a symmetric distribution of fuel pellet cracks and symmetric interfacial slippage, to show that μ < 0.5 if it is assumed that the average hoop stress within the cladding attains yield levels. It was therefore suggested that a high interfacial friction coefficient is unlikely to be operative during a power ramp; this result was used to support the view that interfacial friction effects do not play a dominant role in stress corrosion crack formation within the cladding. (orig.)

  4. A Study on the Structural Integrity Issues of a Dual-Cooled Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung-Kyu; Lee, Kang-Hee; Lee, Young-Ho; Yoon, Kyung-Ho; Kim, Jae-Yong; Song, Kun-Woo [Korea Atomic Energy Research Institute, 1045 Daedeokdaero Yuseong Daejeon 305-353 (Korea, Republic of)

    2009-06-15

    A dual-cooled fuel rod has an internal coolant flow passage in addition to the external one. A remarkable power up-rate can be achieved due to the increased surface area, which may draw great interests from the fuel researchers, designers and vendors. However, it requires effective resolution to the difficult technical issues when a fuel assembly is to be realized. It becomes much more difficult if a tough boundary condition needs to be satisfied such as a compatibility with the existing reactor internal structures. This kind of challenge is tackled through a national R and D project in Korea: to develop the structural components of a dual-cooled fuel that should be compatible with the current OPR 1000 (Korea Standard Nuclear Power Plant) internal structures. Fuel rod supporting structures, top and bottom end pieces and guide tubes are the components. Besides, the fuel rod components have to be developed as well since the fuel rod's geometry becomes much different from the conventional rod's one. The dimension change may well affect the above mentioned structural components. As a part of the work, structural integrity of the components of a dual-cooled fuel rod is studied in this paper. The investigated topics are: i) the thickness determination of a cladding tube (especially outer tube of a large diameter), ii) vibration issue of an inner cladding tube, iii) design concern of plenum spring and spacer. The cladding thickness issue arises due to the increased outside diameter of a fuel rod, which is caused by an internal flow passage formation. Among the criteria for the thickness determination, an elastic buckling criteria was focused on. Theoretical background for the well-known formula (such as a stability problem) was revisited. Verification tests were carried out independently with using a cladding tube of PHWR fuel rod. Results showed that the formula was not conservative to apply for the cladding thickness determination. Minimum thickness for the

  5. Experimental characterization of cooled EGR in a gasoline direct injection engine for reducing fuel consumption and nitrogen oxide emission

    Science.gov (United States)

    Park, Sang-Ki; Lee, Jungkoo; Kim, Kyungcheol; Park, Seongho; Kim, Hyung-Man

    2015-11-01

    The emphasis on increasing fuel economy and reducing emissions is increasing. Attention has turned to how the performance of a gasoline direct injection (GDI) engine can be improved to achieve lower fuel consumption and NOx emission. Therefore, positive effects can reduce fuel consumption and NOx emission as well as knock suppression. The cooled exhaust gas recirculation (EGR) ranges within the characteristic map are characterized from the experimental results at various speeds and brake mean effective pressures in a GDI engine. The results show that the application of cooled EGR system brought in 3.63 % reduction as for the fuel consumption and 4.34 % as for NOx emission.

  6. Gas-Cooled Thorium Reactor with Fuel Block of the Unified Design

    Directory of Open Access Journals (Sweden)

    Igor Shamanin

    2015-01-01

    Full Text Available Scientific researches of new technological platform realization carried out in Russia are based on ideas of nuclear fuel breeding in closed fuel cycle and physical principles of fast neutron reactors. Innovative projects of low-power reactor systems correspond to the new technological platform. High-temperature gas-cooled thorium reactors with good transportability properties, small installation time, and operation without overloading for a long time are considered perspective. Such small modular reactor systems at good commercial, competitive level are capable of creating the basis of the regional power industry of the Russian Federation. The analysis of information about application of thorium as fuel in reactor systems and its perspective use is presented in the work. The results of the first stage of neutron-physical researches of a 3D model of the high-temperature gas-cooled thorium reactor based on the fuel block of the unified design are given. The calculation 3D model for the program code of MCU-5 series was developed. According to the comparison results of neutron-physical characteristics, several optimum reactor core compositions were chosen. The results of calculations of the reactivity margins, neutron flux distribution, and power density in the reactor core for the chosen core compositions are presented in the work.

  7. MELCOR Modeling of Air-Cooled PWR Spent Fuel Assemblies in Water empty Fuel Pools

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Lopez, C.

    2013-07-01

    The OECD Spent Fuel Project (SFP) investigated fuel degradation in case of a complete Loss-Of- Coolant-Accident in a PWR spent fuel pool. Analyses of the SFP PWR ignition tests have been conducted with the 1.86.YT.3084.SFP MELCOR version developed by SNL. The main emphasis has been placed on assessing the MELCOR predictive capability to get reasonable estimates of time-to-ignition and fire front propagation under two configurations: hot neighbor (i.e., adiabatic scenario) and cold neighbor (i.e., heat transfer to adjacent fuel assemblies). A detailed description of hypotheses and approximations adopted in the MELCOR model are provided in the paper. MELCOR results accuracy was notably different between both scenarios. The reasons are highlighted in the paper and based on the results understanding a set of remarks concerning scenarios modeling is given.

  8. Performance of metal and oxide fuels during accidents in a large liquid metal cooled reactor

    International Nuclear Information System (INIS)

    Cahalan, J.; Wigeland, R.; Friedel, G.; Kussmaul, G.; Royl, P.; Moreau, J.; Perks, M.

    1990-01-01

    In a cooperative effort among European and US analysts, an assessment of the comparative safety performance of metal and oxide fuels during accidents in a large (3500 MWt), pool-type, liquid-metal-cooled reactor (LMR) was performed. The study focused on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower (UTOP), and the unprotected loss-of-heat-sink (ULOHS). Emphasis was placed on identification of design features that provide passive, self-limiting responses to upset conditions, and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than oxide-fueled reactors of the same design. 3 refs., 4 figs

  9. Digital Information Platform Design of Fuel Element Engineering For High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Du Yuwei

    2014-01-01

    This product line provide fuel element for high temperature gas-cooled reactor nuclear power plant which is being constructed in Shidao bay in Shandong province. Its annual productive capacity is thirty ten thousands fuel elements whose shape is spherical . Compared with pressurized water fuel , this line has the feature of high radiation .In order to reduce harm to operators, the comprehensive information platform is designed , which can realize integration of automation and management for plant. This platform include two nets, automation net using field bus technique and information net using Ethernet technique ,which realize collection ,control, storage and publish of information.By means of construction, automatization and informatization of product line can reach high level. (author)

  10. Options for treating high-temperature gas-cooled reactor fuel for repository disposal

    Energy Technology Data Exchange (ETDEWEB)

    Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

    1992-02-01

    This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing.

  11. Options for treating high-temperature gas-cooled reactor fuel for repository disposal

    International Nuclear Information System (INIS)

    Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

    1992-02-01

    This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched 235 U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched 235 U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing

  12. Thermal aspects of mixed oxide fuel in application to supercritical water-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grande, L.; Peiman, W.; Rodriguez-Prado, A.; Villamere, B.; Mikhael, S.; Allison, L.; Pioro, I., E-mail: lisa.grande@mycampus.uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

    2010-07-01

    SuperCritical Water-cooled nuclear Reactors (SCWRs) are a renewed technology being developed as one of the Generation IV reactor concepts. This reactor type uses a light water coolant at temperatures and pressures above its critical point. These elevated operating conditions will improve Nuclear Power Plant (NPP) thermal efficiencies by 10 - 15% compared to those of current NPPs. Also, SCWRs will have the ability to utilize a direct cycle, thus decreasing NPP capital and operational costs. The SCWR core has 2 configurations: 1) Pressure Vessel (PV) -type enclosing a fuel assembly and 2) Pressure Tube (PT) -type consisting of individual pressurized channels containing fuel bundles. Canada and Russia are developing PT-type SCWRs. In particular, the Canadian SCWR reactor has an output of 1200 MW{sub el} and will operate at a pressure of 25 MPa with inlet and outlet fuel-channel temperatures of 350 and 625°C, respectively. These extreme operating conditions require alternative fuels and materials to be investigated. Current CANadian Deuterium Uranium (CANDU) nuclear reactor fuel-channel design is based on the use of uranium dioxide (UO{sub 2}) fuel; zirconium alloy sheath (clad) bundle, pressure and calandria tubes. Alternative fuels should be considered to supplement depleting world uranium reserves. This paper studies general thermal aspects of using Mixed OXide (MOX) fuel in an Inconel-600 sheath in a generic PT-type SCWR. The bulk fluid, sheath and fuel centerline temperatures along with the Heat Transfer Coefficient (HTC) profiles were calculated at uniform and non-uniform Axial Heat Flux Profiles (AHFPs). (author)

  13. Compact modeling of a telecom back-up unit powered by air-cooled proton exchange membrane fuel cell

    DEFF Research Database (Denmark)

    Gao, Xin; Kær, Søren Knudsen

    2018-01-01

    Applications of proton exchange membrane fuel cells (PEMFC’s) are expanding in portable, automotive and stationary markets. One promising application is the back-up power for telecommunication applications in remote areas where usually air-cooled PMEFC’s are used. An air-cooled PEMFC system is much...

  14. Analysis of the thorium inclusion in the fuel of a fast reactor cooled by lead

    International Nuclear Information System (INIS)

    Juarez M, L. C.; Francois L, J. L.

    2017-09-01

    In the present work, we first verified a model of the European reactor cooled with lead (ELFR). The calculations were made with the code Monte Carlo serpent 2.27 and the library of cross sections Jeff-3.1. For this verification, three neutron parameters were compared: the evolution of the neutron multiplication factor, the Doppler constant and the effect of the vacuum fraction of the refrigerant, obtaining a good approximation with the reference values. Subsequently, the inclusion of thorium as a fertile material within the fuel was analyzed and the same neutron parameters were compared with the original fuel. The evolution of criticality for the case of thorium fuel differs significantly with respect to that of the original fuel (without thorium); this is due mainly to the breeding of the fissile isotope 233 U. Therefore, is possible to have a longer fuel cycle, favoring the availability factor of the plant, without compromising the performance of the reactor since both the Doppler constant and the effect of the vacuum fraction of the refrigerant show a similar tendency to those of the original fuel, being negative in both cases. (Author)

  15. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Yoon, K. H.; Lee, C. B.

    2014-01-01

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness

  16. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly

    Directory of Open Access Journals (Sweden)

    Ben De Pauw

    2016-04-01

    Full Text Available Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation.

  17. Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

    International Nuclear Information System (INIS)

    Hong, S. G.; Kim, Y. H.; Venneri, F.

    2010-01-01

    In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH 1.8 ) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

  18. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

  19. A Development of Technical Specification of a Research Reactor with Plate Fuels Cooled by Upward Flow

    International Nuclear Information System (INIS)

    Park, Sujin; Kim, Jeongeun; Kim, Hyeonil

    2016-01-01

    The contents of the TS(Technical Specifications) are definitions, safety limits, limiting safety system settings, limiting conditions for operation, surveillance requirements, design features, and administrative controls. TS for Nuclear Power Plants (NPPs) have been developed since many years until now. On the other hands, there are no applicable modernized references of TS for research reactors with many differences from NPPs in purpose and characteristics. Fuel temperature and Departure from Nuclear Boiling Ratio (DNBR) are being used as references from the thermal-hydraulic analysis point of view for determining whether the design of research reactors satisfies acceptance criteria for the nuclear safety or not. Especially for research reactors using plate-type fuels, fuel temperature and critical heat flux, however, are very difficult to measure during the reactor operation. This paper described the outline of main contents of a TS for open-pool research reactor with plate-type fuels using core cooling through passive systems, where acceptance criteria for nuclear safety such as CHF and fuel temperature cannot be directly measured, different from circumstances in NPPs. Thus, three independent variables instead of non-measurable acceptance criteria: fuel temperature and CHF are considered as safety limits, i.e., power, flow, and flow temperature

  20. Definition of breeding gain for the closed fuel cycle and application to a gas cooled fast reactor

    International Nuclear Information System (INIS)

    Van Rooijen, W. F. G.; Kloosterman, J. L.; Van Der Hagen, T. H. J. J.; Van Dam, H.

    2006-01-01

    In this paper a definition is given for the Breeding Gain (BG) of a nuclear reactor, taking into account compositional changes of the fuel during irradiation, cool down and reprocessing. A definition is given for the reactivity weights required to calculate BG. To calculate the effects of changes in the initial fuel composition on BG, first order nuclide perturbation theory is used. The theory is applied to the fuel cycle of GFR600, a 600 MWth Generation IV Gas Cooled Fast Reactor. This reactor should have a closed fuel cycle, with a BG equal to zero, breeding just enough new fuel during irradiation to allow refueling by only adding fertile material. All Heavy Metal is recycled in the closed fuel cycle. The result is that a closed fuel cycle is possible if the reprocessing has low losses ( 238 U, 15% Pu, and low amounts of the Minor Actinides. (authors)

  1. Study and Evaluation of Innovative Fuel Handling Systems for Sodium-Cooled Fast Reactors: Fuel Handling Route Optimization

    Directory of Open Access Journals (Sweden)

    Franck Dechelette

    2014-01-01

    Full Text Available The research for technological improvement and innovation in sodium-cooled fast reactor is a matter of concern in fuel handling systems in a view to perform a better load factor of the reactor thanks to a quicker fuelling/defueling process. An optimized fuel handling route will also limit its investment cost. In that field, CEA has engaged some innovation study either of complete FHR or on the optimization of some specific components. This paper presents the study of three SFR fuel handling route fully described and compared to a reference FHR option. In those three FHR, two use a gas corridor to transfer spent and fresh fuel assembly and the third uses two casks with a sodium pot to evacuate and load an assembly in parallel. All of them are designed for the ASTRID reactor (1500 MWth but can be extrapolated to power reactors and are compatible with the mutualisation of one FHS coupled with two reactors. These three concepts are then intercompared and evaluated with the reference FHR according to four criteria: performances, risk assessment, investment cost, and qualification time. This analysis reveals that the “mixed way” FHR presents interesting solutions mainly in terms of design simplicity and time reduction. Therefore its study will be pursued for ASTRID as an alternative option.

  2. K-infinite trends with burnup, enrichment, and cooling time for BWR fuel assemblies

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1998-08-01

    This report documents the work performed by ORNL for the Yucca Mountain project (YMP) M and O contractor, Framatome Cogema Fuels. The goal of this work was to obtain k inf values for infinite arrays of flooded boiling-water-reactor (BWR) fuel assemblies as a function of various burnup/enrichment and cooling-time combinations. These scenarios simulate expected limiting criticality loading conditions (for a given assembly type) for drift emplacements in a repository. Upon consultation with the YMP staff, a Quad Cities BWR fuel assembly was selected as a baseline assembly. This design consists of seven axial enrichment zones, three of which contain natural uranium oxide. No attempt was made to find a bounding or even typical assembly design due to the wide variety in fuel assembly designs necessary for consideration. The current work concentrates on establishing a baseline analysis, along with a small number of sensitivity studies which can be expected later if desired. As a result of similar studies of this nature, several effects are known to be important in the determination of the final k inf for spent fuel in a cask-like geometry. For a given enrichment there is an optimal burnup: for lower burnups, excess energy (and corresponding excess reactivity) is present in the fuel assembly; for larger burnups, the assembly is overburned and essentially driven by neighboring fuel assemblies. The majority of the burnup/enrichment scenarios included in this study were for some near-optimum burnup/enrichment combinations as determined from Energy Information Administration (EIA) data. Several calculations were performed for under- and over-burned fuel to show these effects

  3. Hybrid heat pipe based passive cooling device for spent nuclear fuel dry storage cask

    International Nuclear Information System (INIS)

    Jeong, Yeong Shin; Bang, In Cheol

    2016-01-01

    Highlights: • Hybrid heat pipe was presented as a passive cooling device for dry storage cask of SNF. • A method to utilize waste heat from spent fuel was suggested using hybrid heat pipe. • CFD analysis was performed to evaluate the thermal performance of hybrid heat pipe. • Hybrid heat pipe can increase safety margin and storage capacity of the dry storage cask. - Abstract: Conventional dry storage facilities for spent nuclear fuel (SNF) were designed to remove decay heat through the natural convection of air, but this method has limited cooling capacity and a possible re-criticality accident in case of flooding. To enhance the safety and capacity of dry storage cask of SNF, hybrid heat pipe-based passive cooling device was suggested. Heat pipe is an excellent passive heat transfer device using the principles of both conduction and phase change of the working fluid. The heat pipe containing neutron absorber material, the so-called hybrid heat pipe, is expected to prevent the re-criticality accidents of SNF and to increase the safety margin during interim and long term storage period. Moreover, a hybrid heat pipe with thermoelectric module, a Stirling engine and a phase change material tank can be used for utilization of the waste heat as heat-transfer medium. Located at the guide tube or instrumentation tube, hybrid heat pipe can remove decay heat from inside the sealed metal cask to outside, decreasing fuel rod temperature. In this paper, a 2-step analysis was performed using computational fluid dynamics code to evaluate the heat and fluid flow inside a cask, which consisted of a single spent fuel assembly simulation and a full-scope dry cask simulation. For a normal dry storage cask, the maximum fuel temperature is 290.0 °C. With hybrid heat pipe cooling, the temperature decreased to 261.6 °C with application of one hybrid heat pipe per assembly, and to 195.1 °C with the application of five hybrid heat pipes per assembly. Therefore, a dry

  4. Thermal hydraulic analysis of Pb-Bi cooled HYPER fuel assemblies using SLTHEN code

    International Nuclear Information System (INIS)

    Tak, Nam Il; Song, Tae Y.; Park, Won S.; Kim, Chang Hyun

    2002-12-01

    In the present work, the existing SLTHEN code, which had been originally developed for subchannel analysis of sodium cooled fast reactors, was modified and applied to the Pb-Bi cooled HYPER core which consists of 237 fuel assemblies (TRU assemblies). In the analysis of single fuel assembly having chopped cosine power profile, the validation and the assessment of usefulness of the modified SLTHEN were focused. In the quantitative comparison, the results of the modified SLTHEN agreed well with those of analytical calculations and of MATRA. For the qualitative approaches, the sensitivity calculations for intra-assembly gap flow and turbulent mixing parameter were used. The sensitivity analysis results showed that the modified SLTHEN can provide reasonable simulations of subchannel thermal hydraulics. In particular, turbulent mixing parameter which is known as the most uncertain parameter in subchannel analyses did not affect largely the maximum cladding temperature. Therefore, it can be said that the results of single assembly show the usefulness of the modified SLTHEN code for thermal hydraulic analysis and design of HYPER under the conceptual design stage. In order to assess intra-assembly heat transfer, subchannel analyses were implemented for two types of 7 assemblies; 1) artificial 7 fuel assemblies to maximize intra-assembly heat transfer, 2) central 7 fuel assemblies in the HYPER reference core. The results showed that the modified SLTHEN can reasonably simulate intra-heat transfer and the amount of intra-assembly heat transfer is not so large in HYPER conditions. Particularly, intra-heat transfer did not affect the maximum coolant and the maximum cladding temperatures which are major parameters in conceptual core designs. The capability of full core thermal hydraulic analysis was confirmed by the analysis of 45 fuel assemblies in 1/6 HYPER core at the first cycle. The SLTHEN predicted that the reference design parameters are acceptable in terms of thermal

  5. Experimental study on heat pipe heat removal capacity for passive cooling of spent fuel pool

    International Nuclear Information System (INIS)

    Xiong, Zhenqin; Wang, Minglu; Gu, Hanyang; Ye, Cheng

    2015-01-01

    Highlights: • A passively cooling SFP heat pipe with an 8.2 m high evaporator was tested. • Heat removed by the heat pipe is in the range of 3.1–16.8 kW. • The heat transfer coefficient of the evaporator is 214–414 W/m 2 /K. • The heat pipe performance is sensitive to the hot water temperature. - Abstract: A loop-type heat pipe system uses natural flow with no electrically driven components. Therefore, such a system was proposed to passively cool spent fuel pools during accidents to improve nuclear power station safety especially for station blackouts such as those in Fukushima. The heat pipe used for a spent fuel pool is large due to the spent fuel pool size. An experimental heat pipe test loop was developed to estimate its heat removal capacity from the spent fuel pool during an accident. The 7.6 m high evaporator is heated by hot water flowing vertically down in an assistant tube with a 207-mm inner diameter. R134a was used as the potential heat pipe working fluid. The liquid R134a level was 3.6 m. The tests were performed for water velocities from 0.7 to 2.1 × 10 −2 m/s with water temperatures from 50 to 90 °C and air velocities from 0.5 m/s to 2.5 m/s. The results indicate significant heat is removed by the heat pipe under conditions that may occur in the spent fuel pool

  6. Sub-channel analysis of LBE-cooled fuel assemblies of accelerator driven systems

    International Nuclear Information System (INIS)

    Cheng, X.; Hwang, D.H.

    2005-01-01

    In the frame of the European PDS-XADS project, two concepts of the sub-critical reactor core cooled by liquid lead-bismuth eutectic (LBE) were proposed. In this paper, the local thermal-hydraulic behavior of both LBE-cooled fuel assemblies was analyzed. For this purpose, the sub-channel analysis code MATRA was selected, and modification was made for its applications to XADS conditions. Compared to the small core concept, the large core concept has a much lower temperatures of coolant, cladding and fuel pins. This enables a short-term realization of the core design using available technologies. The high power density of the small core results in high local temperatures of coolant, cladding and fuel. Both coolant velocity and cladding temperature are such that special attention has to be paid to avoid corrosion and erosion damage of cladding materials. A parametric study shows that under the parameters considered, mixing coefficient has the biggest effect on the coolant temperature distribution, whereas the cladding temperature is strongly affected by the selection of heat transfer correlations. (author)

  7. Achieving salt-cooled reactor goals: economics, variable electricity, no major fuel failures - 15118

    International Nuclear Information System (INIS)

    Forsberg, C.

    2015-01-01

    The Fluoride-salt-cooled High-temperature Reactor (FHR) with a Nuclear air-Brayton Combined Cycle (NACC) and Firebrick Resistance-Heated Energy Storage (FIRES) is a new reactor concept. The FHR uses High-Temperature Gas-cooled Reactor (HTGR) coated-particle fuel and liquid-salt coolants originally developed for molten salt reactors (MSRs) where the fuel was dissolved in the coolant. The FIRES system consists of high-temperature firebrick heated to high temperatures with electricity at times of low electric prices. For a modular FHR operating with a base-load 100 MWe output, the station output can vary from -242 MWe to +242 MWe. The FHR can be built in different sizes. The reactor concept was developed using a top-down approach: markets, requirements, reactor design. The goals are: (1) increase plant revenue by 50 to 100% relative to base-load nuclear plants with capital costs similar to light-water reactors, (2) enable a zero-carbon nuclear renewable electricity grid, and (3) no potential for major fuel failure and thus no potential for major radionuclide offsite releases in a beyond-design-basis accident (BDBA). The basis for the goals and how they may be achieved is described

  8. Gas cooled fast reactor materials: compatibility and reaction kinetics of fuel/matrices couples

    International Nuclear Information System (INIS)

    Lechelle, J.; Aufore, L.; Basini, V.; Belin, R.; Vaudez, S.

    2004-01-01

    Fourth Generation Gas cooled Fast Reactor concept implies a fast neutron spectrum and aims to lead to an iso-generation of minor actinides. Criteria have been defined for these fuels such as: high core filling factor, efficient fuel cooling, low operation temperature, i.e. 400-850 deg C, good fission product retention, burn-ups in the range of 5-8 atom%, Pu content in the range of 15-25%. Materials matching this demand are considered: mixed uranium - plutonium nitrides and carbides as fuels, whereas TiN, TiC, ZrN, ZrC, SiC are investigated as inert matrices. Thermo-chemical compatibility studies have been carried out, mostly for (U,Pu)N/SiC and (U,Pu)N/TiN couples. They have been associated to matching diffusional studies. For the first studies, accidental reactor conditions have been chosen (1600 deg C) so as to select a couple. Results are presented in terms of nature and quantity of resulting phases identified by XRD and SEM for thermodynamical equilibrium experiments. (authors)

  9. Cooling system and climate control of fuel cell electric vehicle (FCEV)

    Energy Technology Data Exchange (ETDEWEB)

    Ap, N.S. [Valeo Engine Cooling, La Varriere (France); Cloarec, M.; Rouveyre, L. [PSA-Renault, Trappes (France)

    2000-07-01

    This paper described the special thermal aspects of the fuel cell electric vehicle (FCEV) program established in 1999 by the combined efforts of the two French car manufacturers PSA and Renault. One of the objectives of the program was to examine the climate control and particularly the air conditioning in the passenger compartment which had not been previously studied. The heat dissipation of FCEV is in the order of 2.5 to 3 times higher than that of a comparable internal combustion engine vehicle (ICEV). In addition, the fuel cell powertrain has two temperature levels. The first level is high for the fuel cell stack and the second is low for the electrical, electronic components and other auxiliaries. This paper presented and described each component of two cooling loops along with the heat performance of each type. The first cooling loop used de-ionized water as a coolant, and the second made use of an ethylene-glycol-water mixture as a coolant. The air conditioning capability is a major aspect of the FCEV thermal management. The electrical source availability creates the condition to introduce an enhanced comfort level. Both winter preheating and summer precooling are possible. refs., figs.

  10. Improvements in quality of as-manufactured fuels for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Minato, Kazuo; Kikuchi, Hironobu; Tobita, Tsutomu; Fukuda, Kousaku; Kaneko, Mitsunobu; Suzuki, Nobuyuki; Yoshimuta, Shigeharu; Tomimoto, Hiroshi.

    1997-01-01

    The mechanisms of coating failure of the fuel particles for the high-temperature gas-cooled reactors during coating and compaction processes of the fuel fabrication were studied to determine a way to reduce the defective particle fraction of the as-manufactured fuels. Through the observation of the defective particles, it was found that the coating failure during the coating process was mainly caused by the strong mechanical shocks to the particles given by violent particle fluidization in the coater and by unloading and loading of the particles. The coating failure during the compaction process was probably related to the direct contact with neighboring particles in the fuel compacts. The coating process was improved by optimizing the mode of the particle fluidization and by developing the process without unloading and loading of the particles at intermediate coating process. The compaction process was improved by optimizing the combination of the pressing temperature and the pressing speed of the overcoated particles. Through these modifications of the fabrication process, the quality of the as-manufactured fuel compacts was improved outstandingly. (author)

  11. Mechanical Performance Evaluation of a Top End Piece for Dual Cooled Fuels

    International Nuclear Information System (INIS)

    Kim, Jae Yong; Yoon, Kyung Ho; Kim, Hyung Kyu; Choi, Woo Seok

    2011-01-01

    A fuel assembly consists of five major components, i.e., a top end piece (TEP), a bottom end piece (BEP), spacer grids (SGs), guide tubes (GTs) and an instrumentation tube (IT): in addition, it also includes fuel rods (FRs). The TEP/BEP should satisfy stress intensity limits according to the conditions A and B of ASME, Section III, Division 1.Subsection NB. In a dual-cooled fuel assembly, the array and position of fuel rods are different from those in a conventional PWR fuel assembly; these changes are necessary for achieving power uprating. The flow plates of the TEP and BEP have to be modified accordingly. The pattern and shape of the flow holes were newly designed. To verify the strength compatibility, the Tresca stress limit according to the ASME code was investigated in the case of an axial load of 22.241 kN. In this paper, the stress linearization procedure for strength evaluation of a newly designed TEP is presented

  12. Lead-Cooled Fast Reactor Systems and the Fuels and Materials Challenges

    Directory of Open Access Journals (Sweden)

    T. R. Allen

    2007-01-01

    Full Text Available Anticipated developments in the consumer energy market have led developers of nuclear energy concepts to consider how innovations in energy technology can be adapted to meet consumer needs. Properties of molten lead or lead-bismuth alloy coolants in lead-cooled fast reactor (LFR systems offer potential advantages for reactors with passive safety characteristics, modular deployment, and fuel cycle flexibility. In addition to realizing those engineering objectives, the feasibility of such systems will rest on development or selection of fuels and materials suitable for use with corrosive lead or lead-bismuth. Three proposed LFR systems, with varying levels of concept maturity, are described to illustrate their associated fuels and materials challenges. Nitride fuels are generally favored for LFR use over metal or oxide fuels due to their compatibility with molten lead and lead-bismuth, in addition to their high atomic density and thermal conductivity. Ferritic/martensitic stainless steels, perhaps with silicon and/or oxide-dispersion additions for enhanced coolant compatibility and improved high-temperature strength, might prove sufficient for low-to-moderate-temperature LFRs, but it appears that ceramics or refractory metal alloys will be necessary for higher-temperature LFR systems intended for production of hydrogen energy carriers.

  13. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering; Greenspan, Ehud [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering

    2015-02-09

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X

  14. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    International Nuclear Information System (INIS)

    Peterson, Per; Greenspan, Ehud

    2015-01-01

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3 . This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel

  15. Investigation of an Alternative Fuel Form for the Liquid Salt Cooled Very High Temperature Reactor (LS-VHTR)

    International Nuclear Information System (INIS)

    Casino, William A. Jr.

    2006-01-01

    Much of the recent studies investigating the use of liquid salts as reactor coolants have utilized a core configuration of graphite prismatic fuel block assemblies with TRISO particles embedded into cylindrical fuel compacts arranged in a triangular pitch lattice. Although many calculations have been performed for this fuel form in gas cooled reactors, it would be instructive to investigate whether an alternative fuel form may yield improved performance for the liquid salt-cooled Very High Temperature Reactor (LS-VHTR). This study investigates how variations in the fuel form will impact the performance of the LS-VHTR during normal and accident conditions and compares the results with a similar analysis that was recently completed for a LS-VHTR core made up of prismatic block fuel. (author)

  16. Strength analysis of fast gas cooled reactor fuel element in conditions of fuel-cladding interraction and non-uniform azimuthal heating

    International Nuclear Information System (INIS)

    Kulikov, I.S.; Tverkovkin, B.E.

    1984-01-01

    The technique and the PRORT mathematical program in FORTRAN language for determining mechanical properties of a fuel element with motionless fuel-cladding interaction taking into account circular temperature non-uniformity in gas-cooled fast reactor conditions are proposed. The calculation results of the fuel element of dissociating gas cooled fast reactor are presented for seven cross-sections over the height of the core. The obtained data testify to appreciable swelling of Cr16Ni15Mo3Nb steel fuel cladding in the conditions of dissociating gas cooled fast reactor through the allowance for the effect of stresses on this essential parameter shows, that its value is lower in comparison with swelling, wherein stresses are not taken into account

  17. Thermal analyses for the spend fuel pool of Taiwan BWR plants during the loss of cooling accident

    Energy Technology Data Exchange (ETDEWEB)

    Chen, B-Y.; Yeh, C-L.; Wei, W-C.; Chen, Y-S., E-mail: onepicemine@iner.gov.tw, E-mail: clinyeh@iner.gov.tw, E-mail: hn150456@iner.gov.tw, E-mail: yschen@iner.gov.tw [Inst. of Nuclear Energy Research, Longtan Township, Taoyuan County, Taiwan (China)

    2014-07-01

    After the Fukushima nuclear accident, the safety of the spent fuel pool has become an important concern. In this study, thermal analysis of the spent fuel pool under a loss of cooling accident is performed. The BWR spent fuel pools in Taiwan are investigated, including the Chinshan, Kuosheng, and Lungmen plants. The transient pool temperature and level behaviors are calculated based on lumped energy balance. After the pool level drops below the top of the fuel, the peak cladding temperature is predicted by the Computational Fluid Dynamics (CFD) analysis. The influence to the cladding temperature of the uniform and checkboard fuel loading patterns is also investigated. (author)

  18. Development of a CVD silica coating for UK advanced gas-cooled nuclear reactor fuel pins

    International Nuclear Information System (INIS)

    Bennett, M.J.; Houlton, M.R.; Moore, D.A.; Foster, A.I.; Swidzinski, M.A.M.

    1983-04-01

    Vapour deposited silica coatings could extend the life of the 20% Cr/25% Ni niobium stabilised (20/25/Nb) stainless steel fuel cladding of the UK advanced gas cooled reactors. A CVD coating process developed originally to be undertaken at atmospheric pressure has now been adapted for operation at reduced pressure. Trials on the LP CVD process have been pursued to the production scale using commercial equipment. The effectiveness of the LP CVD silica coatings in providing protection to 20/25/Nb steel surfaces against oxidation and carbonaceous deposition has been evaluated. (author)

  19. Health and Safety Considerations Associated with Sodium-Cooled Experimental Nuclear Fuel Dismantlement

    Energy Technology Data Exchange (ETDEWEB)

    Carvo, Alan E. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    Between the mid-1970s and the mid-1980s Sandia National Laboratory constructed eleven experimental assemblies to simulate debris beds formed in a sodium-cooled fast breeder reactor. All but one of the assemblies were irradiated. The experimental assemblies were transferred to the Idaho National Laboratory (INL) in 2007 and 2008 for storage, dismantlement, recovery of the uranium for reuse in the nuclear fuel cycle, and disposal of unneeded materials. This paper addresses the effort to dismantle the assemblies down to the primary containment vessel and repackage them for temporary storage until such time as equipment necessary for sodium separation is in place.

  20. Heat transfer from the roughened surface of gas cooled fast breeder reactor fuel element

    International Nuclear Information System (INIS)

    Tang, I.M.

    1979-01-01

    The temperature distributions and the augmentation of heat transfer performance by artificial roughening of a gas cooled fast breeder reactor (GCFR) fuel rod cladding are studied. Numerical solutions are based on the axisymmetric assumption for a two-dimensional model for one rib pitch of axial distance. The local and axial clad temperature distributions are obtained for both the rectangular and ramp rib roughened surface geometries. The transformation of experimentally measured convective heat transfer coefficients, in terms of Stanton number, into GCFR values is studied. In addition, the heat transfer performance of a GCFR fuel rod cladding roughened surface design is evaluated. Approximate analytical solution for correlating an average Stanton number is also obtained and satisfactorily compared with the corresponding numerical result for a GCFR design. The analytical correlation is useful in assessing roughened surface heat transfer performance in scoping studies and conceptual design

  1. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  2. Performance of the diffusion barrier in the metallic fuel in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kim, Jun Hwan; Ryu, Ho Jin; Yang, Seong Woo; Lee, Byoung Oon; Oh, Seok Jin; Lee, Chan Bock; Hahn, Dohee

    2009-01-01

    The objectives in this study are to propose several kinds of barrier materials and to evaluate their performance to prevent a fuel-clad interaction situation between the metallic fuel and the clad material in the Sodium-cooled Fast Reactor (SFR). Metallic foil made from refractory element, electrodeposition of the Cr on the clad surface, and the vapor deposition of the Zr were used as the barrier layers. The diffusion couple test was performed at the temperature of 800degC for 25 hour. The results showed that considerable amount of reaction occurred at the specimen without barrier, whereas excellent performance was observed in that neither reaction nor inter-diffusion occurred in the case of metallic foil made of Cr or V. Electrodeposition was revealed to be excellent provided that optimum deposition condition can be found. Similar to the electro-deposition result, excellent performance observed in the case of vapor deposition condition. (author)

  3. Evaluation of molten fuel containment concepts for gas-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Kang, C.S.; Torri, A.

    1979-10-01

    Four in-vessel molten fuel containment concepts for the GCFR were compared, namely, (1) a ceramic crucible, (2) a borax bath, (3) a heavy metal bath, and (4) a steel bath. The ceramic crucible is the simplest but depends on substantial upward heat removal. The borax bath and the heavy metal bath concepts offer better performance but would require design changes and an increased experimental effort. The steel bath concept is a good compromise and has potential for further improvement by combining it with the essential features of other concepts, i.e., the crucible or the heavy metal bath. It is concluded that several concepts could potentially exploit the normally provided cooled liner barrier in the PCRV cavity for post-accident fuel containment

  4. Analysis of the loss of pool cooling accident in a PWR spent fuel pool with MAAP5

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2014-01-01

    Highlights: • A PWR spent fuel pool was modeled by using MAAP5. • Loss of pool cooling severe accident scenarios were studied. • Loss of pool cooling accidents with two mitigation measures were analyzed. - Abstract: The Fukushima Daiichi nuclear accident shows that it is necessary to study potential severe accidents and corresponding mitigation measures for the spent fuel pool (SFP) of a nuclear power plant (NPP). This paper presents the analysis of loss of pool cooling accident scenarios and the discussion of mitigation measures for the SFP at a pressurized water reactor (PWR) NPP with the MAAP5 code. Analysis of uncompensated loss of water due to the loss of pool cooling with different initial pool water levels of 12.2 m (designated as a reference case) and 10.7 m have been performed based on a MAAP5 input model. Scenarios of the accident such as overheating of uncovered fuel assemblies, oxidation of claddings and hydrogen generation, loss of intactness of fuel rod claddings, and release of radioactive fission products were predicted with the assumption that mitigation measures were unavailable. The results covered a broad spectrum of severe accident evaluations in the SFP. Furthermore, as important mitigation measures, the effects of recovering the SFP cooling system and makeup water in SFP on the accident progressions have also been investigated respectively based on the events of pool water boiling and spent fuels uncovery. Based upon the reference case, three cases with the recovery of SFP cooling system and three other cases with makeup water in SFP have been studied. The results showed that, severe accident might happen if SFP cooling system was not restored timely before the spent fuels started to become uncovered; spent fuels could be completely submerged and severe accident might be avoided if SFP makeup water system provided water with a mass flow rate larger than the average evaporation rate defined as the division of pool water mass above the

  5. Automatic X-ray inspection for escaped coated particles in spherical fuel elements of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Yang, Min; Liu, Qi; Zhao, Hongsheng; Li, Ziqiang; Liu, Bing; Li, Xingdong; Meng, Fanyong

    2014-01-01

    As a core unit of HTGRs (high-temperature gas-cooled reactors), the quality of spherical fuel elements is directly related to the safety and reliability of HTGRs. In line with the design and performance requirements of the spherical fuel elements, no coated fuel particles are permitted to enter the fuel-free zone of a spherical fuel element. For fast and accurate detection of escaped coated fuel particles, X-ray DR (digital radiography) imaging with a step-by-step circular scanning trajectory was adopted for Chinese 10 MW HTGRs. The scanning parameters dominating the volume of the blind zones were optimized to ensure the missing detection of the escaped coated fuel particles is as low as possible. We proposed a dynamic calibration method for tracking the projection of the fuel-free zone accurately, instead of using a fuel-free zone mask of fixed size and position. After the projection data in the fuel-free zone were extracted, image and graphic processing methods were combined for automatic recognition of escaped coated fuel particles, and some practical inspection results were presented. - Highlights: • An X-ray DR imaging system for quality inspection of spherical fuel elements was introduced. • A method for optimizing the blind-zone-related scanning parameter was proposed. • A dynamic calibration method for tracking the fuel-free zone accurately was proposed. • Some inspection results of the disqualified spherical fuel elements with escaped coated fuel particles were presented

  6. Thermochemical Analysis of Gas-Cooled Reactor Fuels Containing Am and Pu Oxides

    International Nuclear Information System (INIS)

    Lindemer, T.B.

    2002-01-01

    Literature values and estimated data for the thermodynamics of the actinide oxides and fission products are applied to explain the chemical behavior in gas-cooled-reactor fuels. Emphasis is placed on the Am-O-C and Pu-O-C systems and the data are used to plot the oxygen chemical potential versus temperature of solid-solid and solid-gas equilibria. These results help explain observations of vaporization in Am oxides, nitrides, and carbides and provide guidance for the ceramic processing of the fuels. The thermodynamic analysis is then extended to the fission product systems and the Si-C-O system. Existing data on oxygen release (primarily as CO) as a function of burnup in the thoria-urania fuel system is reviewed and compared to values calculated from thermodynamic data. The calculations of oxygen release are then extended to the plutonia and americia fuels. Use of ZrC not only as a particle coating that may be more resistant to corrosion by Pd and other noble-metal fission products, but also as a means to getter oxygen released by fission is discussed

  7. Assessment of Gap Conductance Impact on Heat Split in Dual Cooled Annular Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Kun Ho; Chun, Tae Hyun; In, Wang Kee; Yang, Yong Sik; Song, Kun Woo

    2007-07-15

    As a next generation fuel for PWR, a dual cooling annular fuel is being considered promisingly due to various advantage. It is able to increase the thermal margin significantly from not only large heat transfer area but also thin fuel pellet thickness. But the thermal margin at nominal condition could be degraded at certain burnup range because of the inappropriate heat split to inner and outer flow channels. A key factor to influence the heat split is the gap conductances in inner and outer clearances, which varies in terms of thermal expansion, swelling, creep, and so on in the cladding and pellet. As results of the investigation, particularly in the case of low gap conductance when the fuel rod burnup is relatively high, there is high probability that design targets might be violated. Therefore some effort is inevitable to address the concern. But, in parallel, it is necessary to more in detail investigate whether the assumed gap conductance for this analysis and the present design targets are reasonable through further reviews.

  8. Modeling of Rocket Fuel Heating and Cooling Processes in the Interior Receptacle Space of Ground-Based Systems

    Directory of Open Access Journals (Sweden)

    K. I. Denisova

    2016-01-01

    Full Text Available The propellant to fill the fuel tanks of the spacecraft, upper stages, and space rockets on technical and ground-based launch sites before fueling should be prepared to ensure many of its parameters, including temperature, in appropriate condition. Preparation of fuel temperature is arranged through heating and cooling the rocket propellants (RP in the tanks of fueling equipment. Processes of RP temperature preparation are the most energy-intensive and timeconsuming ones, which require that a choice of sustainable technologies and modes of cooling (heating RP provided by the ground-based equipment has been made through modeling of the RP [1] temperature preparation processes at the stage of design and operation of the groundbased fueling equipment.The RP temperature preparation in the tanks of the ground-based systems can be provided through the heat-exchangers built-in the internal space and being external with respect to the tank in which antifreeze, air or liquid nitrogen may be used as the heat transfer media. The papers [1-12], which note a promising use of the liquid nitrogen to cool PR, present schematic diagrams and modeling systems for the RP temperature preparation in the fueling equipment of the ground-based systems.We consider the RP temperature preparation using heat exchangers to be placed directly in RP tanks. Feeding the liquid nitrogen into heat exchanger with the antifreeze provides the cooling mode of PR while a heated air fed there does that of heating. The paper gives the systems of equations and results of modeling the processes of RP temperature preparation, and its estimated efficiency.The systems of equations of cooling and heating RP are derived on the assumption that the heat exchange between the fuel and the antifreeze, as well as between the storage tank and the environment is quasi-stationary.The paper presents calculation results of the fuel temperature in the tank, and coolant temperature in the heat exchanger, as

  9. 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 for ox...

  10. Degradation of H3PO4/PBI High Temperature Polymer Electrolyte Membrane Fuel Cell under Stressed Operating Conditions

    DEFF Research Database (Denmark)

    Zhou, Fan

    performance loss caused by CO poisoning can be alleviated by the presence of water vapor. The CO oxidation via the water gas shift reaction is the main reason for the mitigated CO poisoning with the presence of water vapor. Meanwhile, the CO poisoning can deteriorate with the presence of CO2, although the CO2...... for HT-PEM fuel cell based micro-CHP units for households, the daily startup/shutdown operation is necessary. Moreover, the faults in the H2 supply system or in controlling the reformer can cause the H2 starvation of the HT-PEM fuel cell. The effects of these operating conditions to the degradation...... results in the degradation in cell performance of the HT-PEM fuel cell by increasing the charge transfer resistance and mass transfer resistance. The CO with volume fraction of 1% – 3% can cause significant performance loss to the HT-PEM fuel cell at the operating temperature of 150 oC. The cell...

  11. Fuel research for subcritical and critical GEN-IV systems cooled by heavy liquid metal

    International Nuclear Information System (INIS)

    Sobolev, V.; Verwerft, M.

    2009-01-01

    The participation of the Belgian Nuclear Research Centre SCK-CEN in the worldwide GEN-IV research can be considered as an opportunity. Today's GEN-IV research at SCK-CEN is mainly driven by the interests of the project MYRRHA (Multipurpose hYbrid Research Reactor for High-tech Applications). The main goal of this project is to build at SCK-CEN in Mol a new generation fast spectrum, subcritical, research and materials testing reactor MYRRHA driven by a high-energy proton accelerator. This GEN-IV MTR is cooled by heavy liquid metal (Pb-Bi) and will be used for the ADS concept demonstration, testing and qualification of new fuels, transmutation targets and innovative materials. On the European scale, MYRRHA is integrated in the Euratom FP6 Integrated Project (IP) EUROTRANS (EUROpean research programme for TRANSmutation of high level nuclear waste in an accelerator driven system), as the small-scale experimental machine for transmutation demonstration called XT-ADS. Last but not least, this experimental facility will also demonstrate the technological feasibility of the LFR (Lead-cooled Fast Reactor) GEN-IV concept; in EU the LFR design studies are performed in the framework of the Euratom FP6 ELSY (European Lead-cooled SYstem) project, where SCK-CEN is a partner. Among the research needed to ensure a safe and reliable operation of the MYRRHA/XT ADS reactor, the development and qualification of fuel and cladding materials have been recognized as one of the main key issues to be addressed

  12. Performance evaluation of a stack cooling system using CO{sub 2} air conditioner in fuel cell vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Chul; Won, Jong Phil [Thermal Management Research Center, Korea Automotive Technology Institute, Chungnam 330-912 (Korea); Park, Yong Sun; Lim, Tae Won [Corporate Research and Development Division, Hyundai-Kia Motors, Gyeonggi 449-912 (Korea); Kim, Min Soo [School of Mechanical and Aerospace Engineering, Seoul National University, Seoul 151-744 (Korea)

    2009-01-15

    A relation between the heat release from a fuel cell stack and an air conditioning system's performance was investigated. The air conditioning system installed in a fuel cell vehicle can be used for stack cooling when additional stack heat release is required over a fixed radiator capacity during high vehicle power generation. This study investigated the performance of a stack cooling system using CO{sub 2} air conditioner at various operating conditions. Also, the heat releasing effectiveness and mutual interference were analyzed and compared with those for the conventional radiator cooling system with/without cabin cooling. When the radiator coolant inlet temperature and flow rate were 65 C and 80 L/min, respectively, for the outdoor air inlet speed of 5 m/s, the heat release of the stack cooling system with the aid of CO{sub 2} air conditioner increased up to 36% more than that of the conventional radiator cooling system with cabin cooling. Furthermore, this increased by 7% versus the case without cabin cooling. (author)

  13. The evaluation of minimum cooling period for loading of PWR spent nuclear fuel of a dual purpose metal cask

    Energy Technology Data Exchange (ETDEWEB)

    Dho, Ho Seog; Kim, Tae Man; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

    2016-12-15

    Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R and D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0-4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask.

  14. The evaluation of minimum cooling period for loading of PWR spent nuclear fuel of a dual purpose metal cask

    International Nuclear Information System (INIS)

    Dho, Ho Seog; Kim, Tae Man; Cho, Chun Hyung

    2016-01-01

    Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R and D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0-4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask

  15. Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    Carter, W.L.; Olsen, A.R.

    1979-06-01

    The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing

  16. Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carter, W.L.; Olsen, A.R.

    1979-06-01

    The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.

  17. Development of failed fuel detection and location system in sodium-cooled large reactor. Sampling method of failed fuels under the slit

    International Nuclear Information System (INIS)

    Aizawa, Kousuke; Fujita, Kaoru; Kamide, Hideki; Kasahara, Naoto

    2010-01-01

    A conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR) is in progress as an issue of the 'Fast Reactor Cycle Technology Development (FaCT)' project in Japan. JSFR adopts a Selector-Valve mechanism for the failed fuel detection and location (FFDL) system. The Selector-Valve FFDL system identifies failed fuel subassemblies by sampling sodium from each fuel subassembly outlet and detecting fission product. One of the JSFR design features is employing an upper internal structure (UIS) with a radial slit, in which an arm of fuel handling machine can move and access the fuel assemblies under the UIS. Thus, JSFR cannot place sampling nozzles right above the fuel subassemblies located under the slit. In this study, the sampling method for indentifying under-slit failed fuel subassemblies has been demonstrated by water experiments. (author)

  18. Study on the properties of the fuel compact for High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Lee, Chung-yong; Lee, Sung-yong; Choi, Min-young; Lee, Seung-jae; Jo, Young-ho; Lee, Young-woo; Cho, Moon-sung

    2015-01-01

    High Temperature Gas-cooled Reactors (HTGR), one of the Gen-IV reactors, have been using the fuel element which is manufactured by the graphite matrix, surrounding Tristructural-isotropic (TRISO)-coated Uranium particles. Factors with these characteristics effecting on the matrix of fuel compact are chosen and their impacts on the properties are studied. The fuel elements are considered with two types of concepts for HTGR, which are the block type reactor and the pebble bed reactor. In this paper, the cylinder-formed fuel element for the block type reactor is focused on, which consists of the large part of graphite matrix. One of the most important properties of the graphite matrix is the mechanical strength with the high reliability because the graphite matrix should be enabled to protect the TRISO particles from the irradiation environment and the impact from the outside. In this study, the three kinds of candidate graphites and the two kinds of candidate binder (Phenol and Polyvinyl butyral) were chosen and mixed with each other, formed and heated to measure mechanical properties. The objective of this research is to optimize the materials and composition of the mixture and the forming process by evaluating the mechanical properties before/after carbonization and heat treatment. From the mechanical test results, the mechanical properties of graphite pellets was related to the various conditions such as the contents and kinds of binder, the kinds of graphite and the heat treatments. In the result of the compressive strength and Vicker's hardness, the 10 wt% phenol binder added R+S graphite pellet was relatively higher mechanical properties than other pellets. The contents of Phenol binder, the kinds of graphite powder and the temperature of carbonization and heat treatment are considered important factors for the properties. To optimize the mechanical properties of fuel elements, the role of binders and the properties of graphites will be investigated as

  19. Experience in the development of metal uranium-base nuclear fuel for heavy-water gas-cooled reactors

    International Nuclear Information System (INIS)

    Ashikhmin, V.P.; Vorob'ev, M.A.; Gusarov, M.S.; Davidenko, A.S.; Zelenskij, V.F.; Ivanov, V.E.; Krasnorutskij, V.S.; Petel'guzov, I.A.; Stukalov, A.I.

    1978-01-01

    Investigations were carried out to solve the problem of making the development of radiation-resistant uranium fuel for power reactors including the heavy-water gas-cooled KS-150 reactor. Factors are considered that limit the lifetime of uranium fuel elements, and the ways of suppressing them are discussed. Possible reasons of the insufficient radiation resistance of uranium rod fuel element and the progress attained are analyzed. Some general problems on the fuel manufacture processes are discussed. The main results are presented on the operation of the developed fuel in research reactor loops and the commercial heavy-water KS-150 reactor. The results confirm an exceptionally high radiation resistance of fuel to burn-ups of 1.5-2%. The successful solution of a large number of problems associated with the development of metal uranium fuel provides for new possibilities of using metal uranium in power reactors

  20. Computer simulation of fuel behavior during loss-of-flow accidents in a gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Wehner, T.R.

    1980-01-01

    The sequence of events in a loss-of-flow accident without reactor shutdown in a gas-cooled fast breeder reactor is strongly influenced by the manner in which the fuel deforms. In order to predict the mode of initial gross fuel deformation, welling, melting or cracking, a thermomechanical computer simulation program was developed. Methods and techniques used make the simulation an economical, efficient, and flexible engineering tool. An innovative application of the enthalpy model within a finite difference scheme is used to caculate temperatures in the fuel rod. The method of successive elastic solutions is used to calculate the thermoelastic-creep response. Calculated stresses are compared with a brittle-fracture stress criterion. An independent computer code is used to calculate fission-gas-induced fuel swelling. Results obtained with the computer simulation indicate that swelling is not a mode of initial fuel deformation. Faster transients result in fuel melting, while slower transients result in fuel cracking. For investigated faster coolant flow coastdowns with time constants of 1 second and 10 seconds, compressive stresses in the outer radial portion of the fuel limit fuel swelling and inhibit fuel cracking. For a slower coolant flow coastdown with a 300 second time constant, tensile stresses in the outer radial portion of the fuel induce early fuel cracking before any melting or significant fuel swelling has occurred. Suggestions for further research are discussed. A derived noniterative solution for mechanics calculations may offer an order of magnitude decrease in computational effort

  1. Study of a fuel assembly for the nuclear reactor of IV generation cooled with supercritical water

    International Nuclear Information System (INIS)

    Barragan M, A.; Martin del Campo M, C.; Francois L, J. L.; Espinosa P, G.

    2011-11-01

    In this work a neutron study is presented about a square assembly design of double line of fuel rods, with moderator box to the center of the arrangement, for a nuclear reactor cooled with supercritical water (SCWR). The SCWR reactor was chosen by the characteristics of its design, mainly because is based in light water reactors (PWR and BWR), and the operational experience that has of them allow to use models and similar programs to simulate the fuel and the nucleus of this type of reactors. To develop the necessary models and to carry out the design and analysis of the SCWR reactor, the neutron codes MCNPX and Helios were used. The reason of using both codes, is because the code MCNPX used thoroughly in the neutron simulation of these reactors, it has been our reference code to analyze the results obtained with the Helios code which results are more efficient because its calculation times are minors. In the nucleus design the same parameters for both codes were considered. The results show that the design with Helios is a viable option to simulate these reactors since their values of the neutrons multiplication factor are very similar to those obtained with MCNPX. On the other hand, it could be corroborated that the CASMO-4 code is inadequate to simulate the fuel to the temperature conditions and water pressure in the SCWR. (Author)

  2. Analytical and numerical study on cooling flow field designs performance of PEM fuel cell with variable heat flux

    Science.gov (United States)

    Afshari, Ebrahim; Ziaei-Rad, Masoud; Jahantigh, Nabi

    2016-06-01

    In PEM fuel cells, during electrochemical generation of electricity more than half of the chemical energy of hydrogen is converted to heat. This heat of reactions, if not exhausted properly, would impair the performance and durability of the cell. In general, large scale PEM fuel cells are cooled by liquid water that circulates through coolant flow channels formed in bipolar plates or in dedicated cooling plates. In this paper, a numerical method has been presented to study cooling and temperature distribution of a polymer membrane fuel cell stack. The heat flux on the cooling plate is variable. A three-dimensional model of fluid flow and heat transfer in cooling plates with 15 cm × 15 cm square area is considered and the performances of four different coolant flow field designs, parallel field and serpentine fields are compared in terms of maximum surface temperature, temperature uniformity and pressure drop characteristics. By comparing the results in two cases, the constant and variable heat flux, it is observed that applying constant heat flux instead of variable heat flux which is actually occurring in the fuel cells is not an accurate assumption. The numerical results indicated that the straight flow field model has temperature uniformity index and almost the same temperature difference with the serpentine models, while its pressure drop is less than all of the serpentine models. Another important advantage of this model is the much easier design and building than the spiral models.

  3. Technologies for gas cooled reactor decommissioning, fuel storage and waste disposal. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1998-09-01

    Gas cooled reactors (GCRs) and other graphite moderated reactors have been important part of the world's nuclear programme for the past four decades. The wide diversity in status of this very wide spectrum of plants from initial design to decommissioning was a major consideration of the International Working group on Gas Cooled Reactors which recommended IAEA to convene a Technical Committee Meeting dealing with GCR decommissioning, including spent fuel storage and radiological waste disposal. This Proceedings includes papers 25 papers presented at the Meeting in three sessions entitled: Status of Plant Decommissioning Programmes; Fuels Storage Status and Programmes; waste Disposal and decontamination Practices. Each paper is described here by a separate abstract

  4. Improved lumped models for transient combined convective and radiative cooling of a two-layer spherical fuel element

    International Nuclear Information System (INIS)

    Silva, Alice Cunha da; Su, Jian

    2013-01-01

    The High Temperature Gas cooled Reactor (HTGR) is a fourth generation thermal nuclear reactor, graphite-moderated and helium cooled. The HTGRs have important characteristics making essential the study of these reactors, as well as its fuel element. Examples of these are: high thermal efficiency,low operating costs and construction, passive safety attributes that allow implication of the respective plants. The Pebble Bed Modular Reactor (PBMR) is a HTGR with spherical fuel elements that named the reactor. This fuel element is composed by a particulate region with spherical inclusions, the fuel UO2 particles, dispersed in a graphite matrix and a convective heat transfer by Helium happens on the outer surface of the fuel element. In this work, the transient heat conduction in a spherical fuel element of a pebble-bed high temperature reactor was studied in a transient situation of combined convective and radiative cooling. Improved lumped parameter model was developed for the transient heat conduction in the two-layer composite sphere subjected to combined convective and radiative cooling. The improved lumped model was obtained through two-point Hermite approximations for integrals. Transient combined convective and radiative cooling of the two-layer spherical fuel element was analyzed to illustrate the applicability of the proposed lumped model, with respect to die rent values of the Biot number, the radiation-conduction parameter, the dimensionless thermal contact resistance, the dimensionless inner diameter and coating thickness, and the dimensionless thermal conductivity. It was shown by comparison with numerical solution of the original distributed parameter model that the improved lumped model, with H2,1/H1,1/H0,0 approximation yielded significant improvement of average temperature prediction over the classical lumped model. (author)

  5. Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights September 2010

    International Nuclear Information System (INIS)

    Snead, Lance Lewis; Besmann, Theodore M.; Collins, Emory D.; Bell, Gary L.

    2010-01-01

    The DB Program monthly highlights report for August 2010, ORNL/TM-2010/184, was distributed to program participants by email on September 17. This report discusses: (1) Core and Fuel Analysis - (a) Core Design Optimization in the HTR (high temperature helium-cooled reactor) Prismatic Design (Logos), (b) Core Design Optimization in the HTR Pebble Bed Design (INL), (c) Microfuel analysis for the DB HTR (INL, GA, Logos); (2) Spent Fuel Management - (a) TRISO (tri-structural isotropic) repository behavior (UNLV), (b) Repository performance of TRISO fuel (UCB); (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor) - Synergy with other reactor fuel cycles (GA, Logos); (4) TRU (transuranic elements) HTR Fuel Qualification - (a) Thermochemical Modeling, (b) Actinide and Fission Product Transport, (c) Radiation Damage and Properties; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle - (a) Graphite Recycle (ORNL), (b) Aqueous Reprocessing, (c) Pyrochemical Reprocessing METROX (metal recovery from oxide fuel) Process Development (ANL).

  6. Theoretical investigations on improving performance of cooling systems for fuel cell vehicles; Theoretische Untersuchungen zur Kuehlleistungssteigerung durch innovative Kuehlsysteme fuer Brennstoffzellen-Elektrofahrzeuge

    Energy Technology Data Exchange (ETDEWEB)

    Reichler, Mark

    2008-04-01

    In this work theoretical investigations are carried out for cooling systems, which are used in fuel cell vehicles. This work focuses mainly on the capability of increasing the heat rejection rate by using new alternative cooling systems and by improving the conventional cooling system. Fuel cell vehicles have a higher demand of heat rejection to the ambient than comparable vehicles with combustion engine. The performance of conventional liquid cooling systems, especially at high loads and high ambient temperatures, is often not sufficient anymore. Hence, cooling systems with improved performance are necessary for fuel cell vehicles. The investigations in this work are based on DaimlerChrysler's ''A-Class'' having a PEM-Fuel Cell system integrated. Specific computational models are developed for radiators and condensers to evaluate the performance of different cooling concepts. The models are validated with experimental data. Based on an intensive investigation in the open literature the state of the art of cooling systems for fuel cell vehicles is depicted. Furthermore new cooling concepts as an alternative to the liquid cooling system are presented. The method of cooling the fuel cell by using two-phase transition shows the greatest capability to increase the cooling performance. Hence, this concept is investigated in detail. Two different concepts with three different refrigerants (R113, R245fa und R236fa) are analyzed. Cooling performance of this concept shows improvement of 18.2 up to 32.6 % compared to the conventional liquid cooling system. Thus, a two phase cooling system represents an alternative cooling system for fuel cell vehicles, which should be closer investigated by experiments. (orig.)

  7. Determination of burnup, cooling time and initial enrichment of PWR spent fuel by use of gamma-ray activity ratios

    International Nuclear Information System (INIS)

    Min, D.K.; Park, H.J.; Park, K.J.; Ro, S.G.; Park, H.S.

    1999-01-01

    The Korea Atomic Energy Institute has been developing the algorithms for sequential determination of cooling time, initial enrichment and burnup of the PWR spent fuel assembly by use of gamma ratio measurements, i.e. 134 Cs/ 137 Cs, 154 Eu/ 137 Cs and 106 Ru 137 Cs/( 134 Cs) 2 . Calculations were performed by applying the ORIGEN-S code. This method has advantages over combination techniques of neutron and gamma measurement, because of its simplicity and insensitivity to the measurement geometry. For verifying the algorithms an experiment for determining the cooling time, initial enrichment and burnup of the two PWR spent fuel rods was conducted by use of high-resolution gamma detector (HPGe) system only. This paper describes the method used and interim results of the experiment. This method can be applied for spent fuel characterization, burnup credit and safeguards of the spent fuel management facility

  8. Enthalpy analysis and Heat Exchanger Sizing of an Air-cooled Proton Exchange Membrane Fuel Cell System

    DEFF Research Database (Denmark)

    Gao, Xin; Berning, Torsten; Kær, Søren Knudsen

    below -20 °C in the winter which make liquid-cooled fuel cells impossible. In such cases, air-cooled fuel cell systems are deployed where the air that is fed to the fuel cell serves both as reactant supplier and coolant to remove the waste heat that is generated during fuel cell operation. In some cases...... in order to optimize the operating conditions and the performance of such a system. The adjustable parameters include the fan speed that determines the amount of air that is brought into the system, and the size and rotating speed of the rotating enthalpy wheel. In addition, computational fluid dynamics...... or an ordinary heat exchanger can fulfill the heat recovery demand. Despite the fact that the air enters the stack at a cold temperature, even the forefront of the stack is at a much elevated and desired stack temperature with the help of supplying an acceptable amount of power to an electric stack heater. So...

  9. Safety of the liquid-metal cooled fast breeder reactor and aspects of its fuel cycle

    International Nuclear Information System (INIS)

    Kessler, G.; Papp, R.; Huebel, D.

    1977-01-01

    Design and construction of the sodium-cooled fast reactors KNK-II (20MW(e)) and SNR-300 (300MW(e)) determine the status of safety engineering and safety R and D of LMFBRs in the Federal Republic of Germany. Both prototype fast power reactors have to go through a civil licensing process similar to that applied to present LWRs. A multilevel safety - or defence in depth - approach is applied to the design and construction of fast power reactors. All design data of the fast reactor plant are confirmed by extensive experimental programmes. Design limits of the plant are thoroughly discussed during the licensing process. Important safety R and D programmes have been and are still being performed. A very conservative safety analysis for hypothetical core and other plant accidents is used for present prototype fast reactors. The paper reviews the future trend of development of theoretical methods for accident analysis and the application of experimental results, especially in view of large commercial-type LMFBRs. The safety approach applied to the LMFBR plant is safe operation under normal operating conditions and safe shutdown under off-normal conditions. The consequences of releases of radioactivity to the environment meet the given standards. No chemical reprocessing plant for fast breeder fuel is in operation in the FRG at present; however, R and D work on investigation of all aspects and problem areas of the fast breeder fuel cycle are under way. Systems studies on safety aspects of the fast breeder fuel cycle (transport, reprocessing, fuel fabrication) and its impact on the environment have been performed and the main consequences of these studies are presented in the paper. (author)

  10. High temperature gas-cooled reactors - once-through fuel cycle

    International Nuclear Information System (INIS)

    1979-03-01

    The HTGR, because of a unique combination of design characteristics, is a resource-efficient and cost-effective reactor. In the HTGR, the low power density core, coated particle fuel design, and gas cooling combine to provide high neutron economy, fuel burnup and thermodynamic efficiency. The uranium resource requirements for the current MEU/Th cycle with annual refueling results in a 30-year net U 3 O 8 requirement of 4280 ST/GWe. The basic design of the HTGR refueling scheme, whereby only selected regions of the core need be accessible during each refueling, makes fuel utilization improvements through semi-annual refueling an acceptable alternative in terms of plant availability. This alternative reduces the 30-year U 3 O 8 requirement by about 9%. Additional resource utilization improvements of 10% could be realized by improved fuel management techniques. In addition to improvements achieved in reactor technology, uranium utilization can also be improved by reducing the U-235 content in the depleted uranium (tails) produced by the isotope separation facility. If the Advanced Isotope Separation Technology program, currently under development by the United States, results in a lowering of the tails assay from 0.20 w/o to 0.05 w/o the uranium feed requirement for MEU/Th cycles would be further reduced by 22%. A total improvement of 41% over the already relatively low 4280 ST/GWe net lifetime U 3 O 8 requirement would result in a 2525 ST/GWe 30-year yet U 3 O 8 requirement if all of the potential improvements were realized

  11. Status and prospects for gas cooled reactor fuels. Proceedings of two IAEA meetings held in June 2004 and June 2005

    International Nuclear Information System (INIS)

    2009-04-01

    Recently, efforts to develop high temperature gas cooled reactors with an aim to building futuristic nuclear energy systems with advanced nuclear fuel cycles in the context of the Generation IV International Forum have increased significantly. In addition, several development projects are ongoing, focusing on the burning of weapons grade plutonium, including civil plutonium and other transuranic elements using the 'deep-burn concept', or 'inert matrix fuels', especially in the form of coated particles in gas cooled reactor systems. There is also considerable global interest in developing 'nuclear hydrogen' energy systems using high temperature gas cooled reactors. Apart from these developments, the value of preserving the large technology base developed in Germany, the United Kingdom and the United States of America, as well as information developed in other countries, has also been a subject of interest to the IAEA. At the second annual meeting of the 'technical working group on nuclear fuel cycles options and spent fuel management' (TWG-NFCO), held in Vienna from 28-30 May 2003, it was recommended to hold a technical meeting on Current Status and Future Prospects of Gas Cooled Reactor Fuels. The meeting should cover the technological progress that has been made in the last three years and plan future fabrication and qualification facilities for GCR/HTR fuel. TWG-NFCO considered it timely that this progress should be presented and discussed in the interested community. Recognizing the numerous activities being pursued in many Member States, the IAEA convened the technical meeting on this topic in June 2004 in Vienna. Consequently, an update meeting was held in June 2005, which was hosted by the Kharkov Institute of Physics and Technology of Ukraine to review and integrate the latest developments. This publication combines the results of the technical meeting of June 2004 and the meeting of June 2005. The proceedings presented here contain 25 in depth papers on the

  12. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  13. Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Trellue, Holly Renee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tobin, Stephen Joesph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-19

    The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 and Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.

  14. Control device for can failures of liquid cooled nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Althaus, D.; Mohm, F.; Nyhof, M.

    1974-01-01

    Checking of the fuel or breeding elements of, e.g., sodium-cooled reactors is done by detecting fission products in the coolant, with a flushing gas like argon removing the fission products from the coolant and carrying them to a detector system. In order to increase the escape rate of the fission products from the elements, these are lifted by a hoisting unit into a pit reaching down below the coolant level and are then heated. Heating is achieved by the decay heat or by an additional heating in the receiving pit which at this point is thermically insulated from the exterior. The flushing gas is blown radially into the receiving pit from below. The rising bubbles take along the fission products to a scintillation counter mounted on the head of the receiving vessel. This vessel may have double walls with coolant flowing through the interspace. (DG) [de

  15. Seismic response of high temperature gas-cooled reactor core with block-type fuel, (2)

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1980-01-01

    For the aseismic design of a high temperature gas-cooled reactor (HTGR) with block-type fuel, it is necessary to predict the motion and force of core columns and blocks. To reveal column vibration characteristics in three-dimensional space and impact response, column vibration tests were carried out with a scale model of a one-region section (seven columns) of the HTGR core. The results are as follows: (1) the column has a soft spring characteristic based on stacked blocks connected with loose pins, (2) the column has whirling phenomena, (3) the compression spring force simulating the gas pressure has the effect of raising the column resonance frequency, and (4) the vibration behavior of the stacked block column and impact response of the surrounding columns show agreement between experiment and analysis. (author)

  16. Experimental investigation of pulsating heat pipe performance with regard to fuel cell cooling application

    International Nuclear Information System (INIS)

    Clement, Jason; Wang Xia

    2013-01-01

    A pulsating heat pipe (PHP) is a closed loop, passive heat transfer device. Its operation depends on the phase change of a working fluid within the loop. Design and performance testing of a pulsating heat pipe was conducted under conditions to simulate heat dissipation requirements of a proton exchange membrane (PEM) fuel cell stack. Integration of pulsating heat pipes within bipolar plates of the stack would eliminate the need for ancillary cooling equipment, thus also reducing parasitic losses and increasing energy output. The PHP under investigation, having dimensions of 46.80 cm long and 14.70 cm wide, was constructed from 0.3175 cm copper tube. Heat pipes effectiveness was found to be dependent upon several factors such as energy input, types of working fluid and its filling ratio. Power inputs to the evaporator side of the pulsating heat pipe varied from 80 to 180 W. Working fluids tested included acetone, methanol, and deionized water. Filling ratios between 30 and 70 percent of the total working volume were also examined. Methanol outperformed other fluids tested; with a 45 percent fluid fill ratio and a 120 W power input, the apparatus took the shortest time to reach steady state and had one of the smallest steady state temperature differences. The various conditions studied were chosen to assess the heat pipe's potential as cooling media for PEM fuel cells. - Highlights: ► Methanol as a working fluid outperformed both acetone and water in a pulsating heat pipe. ► Performance for the PHP peaked with methanol and a fill ratio of 45 percent fluid to total volume. ► A smaller resistance was associated with a higher power input to the system.

  17. Development of advanced fabrication technology for high-temperature gas-cooled reactor fuel. Reduction of coating failure fraction

    International Nuclear Information System (INIS)

    Minato, Kazuo; Kikuchi, Hironobu; Fukuda, Kousaku; Tobita, Tsutomu; Yoshimuta, Sigeharu; Suzuki, Nobuyuki; Tomimoto, Hiroshi; Nishimura, Kazuhisa; Oda, Takafumi

    1998-11-01

    The advanced fabrication technology for high-temperature gas-cooled reactor fuel has been developed to reduce the coating failure fraction of the fuel particles, which leads to an improvement of the reactor safety. The present report reviews the results of the relevant work. The mechanisms of the coating failure of the fuel particles during coating and compaction processes of the fuel fabrication were studied to determine a way to reduce the coating failure fraction of the fuel. The coating process was improved by optimizing the mode of the particle fluidization and by developing the process without unloading and loading of the particles at intermediate coating process. The compaction process was improved by optimizing the combination of the pressing temperature and the pressing speed of the overcoated particles. Through these modifications of the fabrication process, the quality of the fuel was improved outstandingly. (author)

  18. An Advanced Sodium-Cooled Fast Reactor Core Concept Using Uranium-Free Metallic Fuels for Maximizing TRU Burning Rate

    Directory of Open Access Journals (Sweden)

    Wuseong You

    2017-12-01

    Full Text Available In this paper, we designed and analyzed advanced sodium-cooled fast reactor cores using uranium-free metallic fuels for maximizing burning rate of transuranics (TRU nuclides from PWR spent fuels. It is well known that the removal of fertile nuclides such as 238U from fuels in liquid metal cooled fast reactor leads to the degradation of important safety parameters such as the Doppler coefficient, coolant void worth, and delayed neutron fraction. To resolve the degradation of the Doppler coefficient, we considered adding resonant nuclides to the uranium-free metallic fuels. The analysis results showed that the cores using uranium-free fuels loaded with tungsten instead of uranium have a significantly lower burnup reactivity swing and more negative Doppler coefficients than the core using uranium-free fuels without resonant nuclides. In addition, we considered the use of axially central B4C absorber region and moderator rods to further improve safety parameters such as sodium void worth, burnup reactivity swing, and the Doppler coefficient. The results of the analysis showed that the final design core can consume ~353 kg per cycle and satisfies self-controllability under unprotected accidents. The fuel cycle analysis showed that the PWR–SFR coupling fuel cycle option drastically reduces the amount of waste going to repository and the SFR burner can consume the amount of TRUs discharged from 3.72 PWRs generating the same electricity.

  19. Onsite nondestructive examination techniques for irradiated water-cooled power reactor fuel

    International Nuclear Information System (INIS)

    1981-03-01

    The International Atomic Energy Agency, in response to the recommendations from several Member States, has prepared this Guidebook on Onsite Non-Destructive Techniques for Irradiated Water-Cooled Power Reactor Fuel with the assistance of a number of experts and organizations in this field. During the preparation of this report it became evident that a comparison between different techniques is a most difficult task and depends on a number of factors related to fuel design, plant characteristics and operating conditions. Consequently the emphasis of the report is on the survey of different techniques presently available. It is also to be noted that because the degree of development for any given technique varies significantly among organizations, it is understood that the report should not be used as consensus standard of the minimum capabilities for each class of techniques, nor does it give recommendations in the regulatory sense. Furthermore, the inclusion of some commercial pieces of equipment, services and other products are for illustrative purposes only and neither implies any preference by the Agency nor can the Agency be liable for any material presented in the report

  20. COUPLED SIMULATION OF GAS COOLED FAST REACTOR FUEL ASSEMBLY WITH NESTLE CODE SYSTEM

    Directory of Open Access Journals (Sweden)

    Filip Osusky

    2018-05-01

    Full Text Available The paper is focused on coupled calculation of the Gas Cooled Fast Reactor. The proper modelling of coupled neutronics and thermal-hydraulics is the corner stone for future safety assessment of the control and emergency systems. Nowadays, the system and channel thermal-hydraulic codes are accepted by the national regulatory authorities in European Union for license purposes, therefore the code NESTLE was used for the simulation. The NESTLE code is a coupled multigroup neutron diffusion code with thermal-hydraulic sub-channel code. In the paper, the validation of NESTLE code 5.2.1 installation is presented. The processing of fuel assembly homogeneous parametric cross-section library for NESTLE code simulation is made by the sequence TRITON of SCALE code package system. The simulated case in the NESTLE code is one fuel assembly of GFR2400 concept with reflective boundary condition in radial direction and zero flux boundary condition in axial direction. The results of coupled calculation are presented and are consistent with the GFR2400 study of the GoFastR project.

  1. Some methods of failed fuel element detection in water cooled reactors

    International Nuclear Information System (INIS)

    Strindehag, O.M.

    1976-01-01

    The methods are surveyed using fission products released in the coolant for the detection of failed fuel elements in water cooled reactors. The classification of the detection methods is made with respect to fission product detection in the coolant and to gaseous fission product detection. The detection systems are listed used for the AGESTA power reactor and for the experimental loops of the RA research reactor based on the detection of either gaseous fission products or gaseous daughter products. The AGESTA reactor detection systems using electrostatic precipitators consist of five precipitator channels of which three are intended for detection and two for localization. A special detection unit was developed for the failed fuel element detection in the R-2 reactor experimental steam loop. Its description is listed. In the reactor pressurized-water loop a Cherenkov counter was used in the detection of fission products. An ion exchange monitor whose application is described was used in the total measurement of the main coolant flow in the AGESTA reactor. (J.P.)

  2. Thermal analysis of lithium cooled natural circulation loop module for fuel rod testing in the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Eyler, L.L.; Kim, D.; Stover, R.L.; Beaver, T.R.

    1987-01-01

    Maximum heat removal capability of a lithium cooled natural circulation fuel rod test module design is determined. Loop geometry is optimized within limitations of design specifications for nominal operation temperatures, materials, and test module environment. Results provide test module operation limits and range of potential uncertainties. 3 refs., 12 figs

  3. Recovery of perchloroethylene scrubbing medium generated in the refabrication of high-temperature gas-cooled reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Judd, M.S.; Van Cleve, J.E. Jr.; Rainey, W.T. Jr.

    1976-11-01

    During the refabrication of high-temperature gas-cooled reactor (HTGR) fuel, perchloroethylene (C/sub 2/Cl/sub 4/) is used as the nonmoderating scrubbing medium to remove condensable hydrocarbons, carbon soot, and uranium-bearing particulates from the off-gas streams. The process by which the contaminated perchloroethylene is recycled is discussed.

  4. Recovery of perchloroethylene scrubbing medium generated in the refabrication of high-temperature gas-cooled reactor fuel

    International Nuclear Information System (INIS)

    Judd, M.S.; Van Cleve, J.E. Jr.; Rainey, W.T. Jr.

    1976-11-01

    During the refabrication of high-temperature gas-cooled reactor (HTGR) fuel, perchloroethylene (C 2 Cl 4 ) is used as the nonmoderating scrubbing medium to remove condensable hydrocarbons, carbon soot, and uranium-bearing particulates from the off-gas streams. The process by which the contaminated perchloroethylene is recycled is discussed

  5. Study on disposal method of graphite blocks and storage of spent fuel for modular gas-cooled reactor. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Sumita, Junya; Sawa, Kazuhiro; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tsuchie, Yasuo; Urakami, Masao [Japan Atomic Power Co., Tokyo (Japan)

    2003-02-01

    This report describes the result of study on disposal method of graphite blocks in future block-type reactor. Present study was carried out within a framework of joint research, ''Research of Modular High Temperature Gas-cooled Reactors (No. 3)'', between Japan Atomic Energy Research Institute (JAERI) and the Japan Atomic Power Company (JAPCO), in 2000. In this study, activities in fuel and reflector graphite blocks were evaluated and were compared with the disposal limits defined as low-level of radioactive waste. As a result, it was found that the activity for only C-14 was higher than disposal limits for the low-level of radioactive waste and that the amount of air in the graphite is important to evaluate precisely of C-14 activity. In addition, spent fuels can be stored in air-cooled condition at least after two years cooling in the storage pool. (author)

  6. The influence of the mechanical properties on fuel rod support characteristics – A case study of dual cooled fuel

    International Nuclear Information System (INIS)

    Kim, Jae Yong; Kim, Hyung Kyu; Ko, Sung Ho

    2014-01-01

    Highlights: • Spring characterization test and analysis were performed to obtain characteristic curves of modified H-type spring. • Using an actual mechanical property is needed to correctly predict the spring characteristics. • The characteristics during unloading should be used for a spacer grid support design. - Abstract: This paper concerns a finite element analysis for a spacer grid support (spring and dimple) design. An accurate prediction of the support characteristics (contact force vs. deflection) is the most crucial in the design by analysis. It is found that the mechanical properties are the key parameter to simulate the characteristics as close as the experimental results after using three different sets of mechanical property data including the actual tensile test results of the present material for a spacer grid of a dual cooled fuel. Besides, the validity of using the characteristics during unloading process is also discussed incorporating a possible overshoot of the support. The coincidence between the present finite element prediction and experimental results is quite good: less than 3.09% at most

  7. Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1. 11, September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

    1982-01-01

    This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts.

  8. Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1.11, September 1981

    International Nuclear Information System (INIS)

    Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

    1982-01-01

    This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts

  9. Optimization of a thermoelectric generator subsystem for high temperature PEM fuel cell exhaust heat recovery

    DEFF Research Database (Denmark)

    Gao, Xin; Andreasen, Søren Juhl; Kær, Søren Knudsen

    2014-01-01

    In previous work, a thermoelectric (TE) exhaust heat recovery subsystem for a high temperature polymer electrolyte membrane (HT-PEM) fuel cell stack was developed and modeled. Numerical simulations were conducted and have identified an optimized subsystem configuration and 4 types of compact heat...... modules are now connected into branches. The procedures of designing and optimizing this TE exhaust heat recovery subsystem are drawn out. The contribution of TE exhaust heat recovery to the HT-PEM fuel cell power system is preliminarily concluded. Its feasibility is also discussed....... exchangers with superior performance for further analysis. In this work, the on-design performances of the 4 heat exchangers are more thoroughly assessed on their corresponding optimized subsystem configurations. Afterward, their off-design performances are compared on the whole working range of the fuel...

  10. Phase characteristics of rare earth elements in metallic fuel for a sodium-cooled fast reactor by injection casting

    Energy Technology Data Exchange (ETDEWEB)

    Kuk, Seoung Woo, E-mail: swkuk@kaeri.re.kr [Next Generation Fuel Development Division, Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon, 34057 (Korea, Republic of); Kim, Ki Hwan; Kim, Jong Hwan; Song, Hoon; Oh, Seok Jin; Park, Jeong-Yong; Lee, Chan Bock [Next Generation Fuel Development Division, Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon, 34057 (Korea, Republic of); Youn, Young-Sang [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon, 34057 (Korea, Republic of); Kim, Jong-Yun [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon, 34057 (Korea, Republic of); Radiochemistry & Nuclear Nonproliferation, University of Science & Technology, Gajeong-ro 217, Yuseong-gu, Daejeon, 34113 (Korea, Republic of)

    2017-04-01

    Uranium-zirconium-rare earth (U-Zr-RE) fuel slugs for a sodium-cooled fast reactor were manufactured using a modified injection casting method, and investigated with respect to their uniformity, distribution, composition, and phase behavior according to RE content. Nd, Ce, Pr, and La were chosen as four representative lanthanide elements because they are considered to be major RE components of fuel ingots after pyroprocessing. Immiscible layers were found on the top layers of the melt-residue commensurate with higher fuel slug RE content. Scanning electron microscopy-energy-dispersive X-ray spectroscopy (SEM-EDS) data showed that RE elements in the melt-residue were distributed uniformly throughout the fuel slugs. RE element agglomeration did not contaminate the fuel slugs but strongly affected the RE content of the slugs.

  11. Study of the strength of the internal can for internally and externally cooled fuel elements intended for gas graphite reactors

    International Nuclear Information System (INIS)

    Boudouresque, B.; Courcon, P.; Lestiboubois, G.

    1964-01-01

    The cartridge of an internally and externally cooled annular fuel element used in gas-graphite reactors is made up of an uranium fuel tube, an external can and an internal can made of magnesium alloy. For the thermal exchange between the internal can and the fuel to be satisfactory, it is necessary for the can to stay in contact with the uranium under all temperature conditions. This report, based on a theoretical study, shows how the internal can fuel gap varies during the processes of canning, charging into the reactor and thermal cycling. The following parameters are considered: tube diameter, pressure of the heat carrying gas, gas entry temperature, plasticity of the can alloy. It is shown that for all operating conditions the internal can of a 77 x 95 element, planned for a gas-graphite reactor with a 40 kg/cm 2 gas pressure, should remain in contact with the fuel. (authors) [fr

  12. All heavy metals closed-cycle analysis on water-cooled reactors of uranium and thorium fuel cycle systems

    International Nuclear Information System (INIS)

    Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Takaki, Naoyuki

    2009-01-01

    Uranium and Thorium fuels as the basis fuel of nuclear energy utilization has been used for several reactor types which produce trans-uranium or trans-thorium as 'by product' nuclear reaction with higher mass number and the remaining uranium and thorium fuels. The utilization of recycled spent fuel as world wide concerns are spent fuel of uranium and plutonium and in some cases using recycled minor actinide (MA). Those fuel schemes are used for improving an optimum nuclear fuel utilization as well to reduce the radioactive waste from spent fuels. A closed-cycle analysis of all heavy metals on water-cooled cases for both uranium and thorium fuel cycles has been investigated to evaluate the criticality condition, breeding performances, uranium or thorium utilization capability and void reactivity condition. Water-cooled reactor is used for the basic design study including light water and heavy water-cooled as an established technology as well as commercialized nuclear technologies. A developed coupling code of equilibrium fuel cycle burnup code and cell calculation of SRAC code are used for optimization analysis with JENDL 3.3 as nuclear data library. An equilibrium burnup calculation is adopted for estimating an equilibrium state condition of nuclide composition and cell calculation is performed for calculating microscopic neutron cross-sections and fluxes in relation to the effect of different fuel compositions, different fuel pin types and moderation ratios. The sensitivity analysis such as criticality, breeding performance, and void reactivity are strongly depends on moderation ratio and each fuel case has its trend as a function of moderation ratio. Heavy water coolant shows better breeding performance compared with light water coolant, however, it obtains less negative or more positive void reactivity. Equilibrium nuclide compositions are also evaluated to show the production of main nuclides and also to analyze the isotopic composition pattern especially

  13. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) with Silicon-Carbide-Matrix Coated-Particle Fuel

    International Nuclear Information System (INIS)

    Forsberg, C. W.; Snead, Lance Lewis; Katoh, Yutai

    2012-01-01

    The FHR is a new reactor concept that uses coated-particle fuel and a low-pressure liquid-salt coolant. Its neutronics are similar to a high-temperature gas-cooled reactor (HTGR). The power density is 5 to 10 times higher because of the superior cooling properties of liquids versus gases. The leading candidate coolant salt is a mixture of 7 LiF and BeF 2 (FLiBe) possessing a boiling point above 1300 C and the figure of merit ρC p (volumetric heat capacity) for the salt slightly superior to water. Studies are underway to define a near-term base-line concept while understanding longer-term options. Near-term options use graphite-matrix coated-particle fuel where the graphite is both a structural component and the primary neutron moderator. It is the same basic fuel used in HTGRs. The fuel can take several geometric forms with a pebble bed being the leading contender. Recent work on silicon-carbide-matrix (SiCm) coated-particle fuel may create a second longer-term fuel option. SiCm coated-particle fuels are currently being investigated for use in light-water reactors. The replacement of the graphite matrix with a SiCm creates a new family of fuels. The first motivation behind the effort is to take advantage of the superior radiation resistance of SiC compared to graphite in order to provide a stable matrix for hosting coated fuel particles. The second motivation is a much more rugged fuel under accident, repository, and other conditions.

  14. Void reactivity decomposition for the Sodium-cooled Fast Reactor in equilibrium fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Sun Kaichao, E-mail: kaichao.sun@psi.ch [Paul Scherrer Institut (PSI), 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), 1015 Lausanne (Switzerland); Krepel, Jiri; Mikityuk, Konstantin; Pelloni, Sandro [Paul Scherrer Institut (PSI), 5232 Villigen PSI (Switzerland); Chawla, Rakesh [Paul Scherrer Institut (PSI), 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), 1015 Lausanne (Switzerland)

    2011-07-15

    Highlights: > We analyze the void reactivity effect for three ESFR core fuel cycle states. > The void reactivity effect is decomposed by neutron balance method. > Novelly, the normalization to the integral flux in the active core is applied. > The decomposition is compared with the perturbation theory based results. > The mechanism and the differences of the void reactivity effect are explained. - Abstract: The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many advantages, but has one dominating neutronic drawback - a positive sodium void reactivity. The aim of this study is to develop and apply a methodology, which should help better understand the causes and consequences of the sodium void effect. It focuses not only on the beginning-of-life (BOL) state of the core, but also on the beginning of open and closed equilibrium (BOC and BEC, respectively) fuel cycle conditions. The deeper understanding of the principal phenomena involved may subsequently lead to appropriate optimization studies. Various voiding scenarios, corresponding to different spatial zones, e.g. node or assembly, have been analyzed, and the most conservative case - the voiding of both inner and outer fuel zones - has been selected as the reference scenario. On the basis of the neutron balance method, the corresponding SFR void reactivity has been decomposed reaction-, isotope-, and energy-group-wise. Complementary results, based on generalized perturbation theory and sensitivity analysis, are also presented. The numerical analysis for both neutron balance and perturbation theory methods has been carried out using appropriate modules of the ERANOS code system. A strong correlation between the flux worth, i.e. the product of flux and adjoint flux, and the void reactivity importance distributions has been found for the node- and assembly-wise voiding scenarios. The neutron balance based decomposition has shown that the void effect is caused mainly by the

  15. Void reactivity decomposition for the Sodium-cooled Fast Reactor in equilibrium fuel cycle

    International Nuclear Information System (INIS)

    Sun Kaichao; Krepel, Jiri; Mikityuk, Konstantin; Pelloni, Sandro; Chawla, Rakesh

    2011-01-01

    Highlights: → We analyze the void reactivity effect for three ESFR core fuel cycle states. → The void reactivity effect is decomposed by neutron balance method. → Novelly, the normalization to the integral flux in the active core is applied. → The decomposition is compared with the perturbation theory based results. → The mechanism and the differences of the void reactivity effect are explained. - Abstract: The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many advantages, but has one dominating neutronic drawback - a positive sodium void reactivity. The aim of this study is to develop and apply a methodology, which should help better understand the causes and consequences of the sodium void effect. It focuses not only on the beginning-of-life (BOL) state of the core, but also on the beginning of open and closed equilibrium (BOC and BEC, respectively) fuel cycle conditions. The deeper understanding of the principal phenomena involved may subsequently lead to appropriate optimization studies. Various voiding scenarios, corresponding to different spatial zones, e.g. node or assembly, have been analyzed, and the most conservative case - the voiding of both inner and outer fuel zones - has been selected as the reference scenario. On the basis of the neutron balance method, the corresponding SFR void reactivity has been decomposed reaction-, isotope-, and energy-group-wise. Complementary results, based on generalized perturbation theory and sensitivity analysis, are also presented. The numerical analysis for both neutron balance and perturbation theory methods has been carried out using appropriate modules of the ERANOS code system. A strong correlation between the flux worth, i.e. the product of flux and adjoint flux, and the void reactivity importance distributions has been found for the node- and assembly-wise voiding scenarios. The neutron balance based decomposition has shown that the void effect is caused mainly

  16. Performance of highly rated UO2 fuel in the WR-1 organic-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schankula, M. H.; Hastings, I. J.

    1977-07-15

    Information on oxide fuel behaviour in organic coolant was required as part of the organic-cooled power reactor (OCR) study. Of major interest were data on the release of fission gases from fuel operating at high fuel surface temperatures and low external restraint; features which are peculiar to the OCR. To provide these and other data, UO2 fuel with cold-worked Zr-2.5wt%Nb sheathing was irradiated in the WR-1 organic-cooled reactor to burnups of 135-154 MWh/kgU at a time-averaged linear power of 60-63 kW/m. Elements with 0.38 and 0.69 mm thick sheathing showed maximum diametral increases averaging 3.7 and 1.7% respectively at pellet mid-planes. Reduced fuel/sheath heat transfer resulting from a difference between internal gas pressure and coolant pressure produced high operating temperatures, and there was evidence of central melting in some elements. Fission gas releases were 30-60%. In the heat affected zone adjacent to brazed appendages, the diametral increases were lower, averaging 0.9 and 0.5% for 0.38 and 0.69 mm thick sheathing respectively. Heat treatment during the brazing process produced a local improvement in sheath creep strength. Highly rated oxide fuel irradiated in organic coolant will require sheathing with improved high temperature creep properties; heat-treated Zr-2.5 wt% Nb may provide this improvement.

  17. Sodium-cooled Fast Reactor Cores using Uranium-Free Metallic Fuels for Maximizing TRU Support Ratio

    International Nuclear Information System (INIS)

    You, WuSeung; Hong, Ser Gi

    2014-01-01

    The depleted uranium plays important roles in the SFR burner cores because it substantially contributes to the inherent safety of the core through the negative Doppler coefficient and large delayed neutron. However, the use of depleted uranium as a diluent nuclide leads to a limited value of TRU support ratio due to the generation of TRUs through the breeding. In this paper, we designed sodium cooled fast reactor (SFR) cores having uranium-free fuels 3,4 for maximization of TRU consumption rate. However, the uranium-free fuelled burner cores can be penalized by unacceptably small values of the Doppler coefficient and small delayed neutron fraction. In this work, metallic fuels of TRU-(W or Ni)-Zr are considered to improve the performances of the uranium-free cores. The objective of this work is to consistently compare the neutronic performances of uranium-free sodium cooled fast reactor cores having TRU-Zr metallic fuels added with Ni or W and also to clarify what are the problematic features to be resolved. In this paper, a consistent comparative study of 400MWe sodium cooled burner cores having uranium-based fuels and uranium-free fuels was done to analyze the relative core neutronic features. Also, we proposed a uranium-free metallic fuel based on Nickel. From the results, it is found that tungsten-based uranium-free metallic fuel gives large negative Doppler coefficient due to high resonance of tungsten isotopes but this core has large sodium void worth and small effective delayed neutron fraction while the nickel-based uranium-free metallic fuelled core has less negative Doppler coefficient but smaller sodium void worth and larger effective delayed neutron fraction than the tungsten-based one. On the other hand, the core having TRU-Zr has very high burnup reactivity swing which may be problematic in compensating it using control rods and the least negative Doppler coefficient

  18. Thermal-hydraulic numerical simulation of fuel sub-assembly for Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Saxena, Aakanksha

    2014-01-01

    The thesis focuses on the numerical simulation of sodium flow in wire wrapped sub-assembly of Sodium-cooled Fast Reactor (SFR). First calculations were carried out by a time averaging approach called RANS (Reynolds- Averaged Navier-Stokes equations) using industrial code STAR-CCM+. This study gives a clear understanding of heat transfer between the fuel pin and sodium. The main variables of the macroscopic flow are in agreement with correlations used hitherto. However, to obtain a detailed description of temperature fluctuations around the spacer wire, more accurate approaches like LES (Large Eddy Simulation) and DNS (Direct Numerical Simulation) are clearly needed. For LES approach, the code TRIO U was used and for the DNS approach, a research code was used. These approaches require a considerable long calculation time which leads to the need of representative but simplified geometry. The DNS approach enables us to study the thermal hydraulics of sodium that has very low Prandtl number inducing a very different behavior of thermal field in comparison to the hydraulic field. The LES approach is used to study the local region of sub-assembly. This study shows that spacer wire generates the local hot spots (∼20 C) on the wake side of spacer wire with respect to the sodium flow at the region of contact with the fuel pin. Temperature fluctuations around the spacer wire are low (∼1 C-2 C). Under nominal operation, the spectral analysis shows the absence of any dominant peak for temperature oscillations at low frequency (2-10 Hz). The obtained spectra of temperature oscillations can be used as an input for further mechanical studies to determine its impact on the solid structures. (author) [fr

  19. Gamma spectrometric characterization of short cooling time nuclear spent fuels using hemispheric CdZnTe detectors

    CERN Document Server

    Lebrun, A; Szabó, J L; Arenas-Carrasco, J; Arlt, R; Dubreuil, A; Esmailpur-Kazerouni, K

    2000-01-01

    After years of cooling, nuclear spent fuel gamma emissions are mainly due to caesium isotopes which are emitters at 605, 662 and 796-801 keV. Extensive work has been done on such fuels using various CdTe or CdZnTe probes. When fuels have to be measured after short cooling time (during NPP outage) the spectrum is much more complex due to the important contributions of niobium and zirconium in the 700 keV range. For the first time in a nuclear power plant, four spent fuels of the Kozloduy VVER reactor no 4 were measured during outage, 37 days after shutdown of the reactor. In such conditions, good resolution is of particular interest, so a 20 mm sup 3 hemispheric crystal was used with a resolution better than 7 keV at 662 keV. This paper presents the experimental device and analyzes the results which show that CdZnTe commercially available detectors enabled us to perform a semi-quantitative determination of the burn-up after a short cooling time. In addition, it is discussed how a burn-up evolution code (CESAR)...

  20. Thermal-hydraulic analysis of bayonet cooling thimble in fuel drain tank of ORNL 10 MW MSRE

    International Nuclear Information System (INIS)

    Sun Lu; Sun Licheng; Yan Changqi

    2012-01-01

    The residual heat removal system of molten salt reactor designed by ORNL, using molten salt as fuel and draining the fuel into fuel drain tank after shutdown of the reactor, removes the decay heat by the circulation of water through the bayonet cooling thimbles in the fuel drain tank. According to structural features of the bayonet cooling thimbles in ORNL 10 MW molten salt reactor experiment (MSRE), this paper presents the analytical results of the influence of the width of gas gap and the width of steam riser on the heat removal ability and the natural circulation of the cooling water, etc. The analysis results show that, when the width of gas gap range from 3.1 mm to 5.1 mm, the change of heat dissipation power and natural circulation flow rate are both less than 5%; when the width of steam riser changes from 3.6 mm to 5.1 mm, the flow mass of the natural circulation change from 1.9 kg/s to 4.79 kg/s, with a slightly effect on the heat transfer efficiency of the system. (authors)

  1. A COMPUTATIONAL FLUID DYNAMICS ANALYSIS OF AIR FLOW THROUGH A TELECOM BACK-UP UNIT POWERED BY AN AIR-COOLED PROTON EXCHANGE MEMBRANE FUEL CELL

    DEFF Research Database (Denmark)

    Gao, Xin; Berning, Torsten; Kær, Søren Knudsen

    2016-01-01

    Proton exchange membrane fuel cells (PEMFC’s) are currently being commercialized for various applications ranging from automotive to stationary such as powering telecom back-up units. In PEMFC’s, oxygen from air is internally combined with hydrogen to form water and produce electricity and heat....... This product heat has to be effectively removed from the fuel cell, and while automotive fuel cells are usually liquid-cooled using a secondary coolant loop similar to the internal combustion engines, stationary fuel cell systems as they are used for telecom back-up applications often rely on excessive air fed...... to the fuel cell cathode to remove the heat. Thereby, the fuel cell system is much simpler and cheaper while the fuel cell performance is substantially lower compared to automotive fuel cells. This work presents a computational fluid dynamics analysis on the heat management of an air-cooled fuel cell powered...

  2. Detection of gas-permeable fuel particles for highl 7490 temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Thiele, B.A.; Stinton, D.P.; Costanzo, D.A.

    1980-01-01

    Fuel for High-Temperature Gas-Cooled Reactors (HTGR) consists of uranium oxide-carbide and thoria microspheres coated with layers of pyrolytic carbon and silicon carbide. The pyrolytic carbon coatings must be gas-tight to perform properly during irradiation. Therefore, particles must be carefully characterized to determine the number of defective particles (ie bare kernels, and cracked or permeable coatings). Although techniques are available to determine the number of bare kernels or cracked coatings, no reliable technique has been available to measure coating permeability. This work describes a technique recently developed to determine whether coatings for a batch of particles are gas-tight or permeable. Although most of this study was performed on Biso-coated particles, the technique applies equally well to Triso-coated particles. About 150 randomly selected Biso-particle batches were studied in this work. These batches were first subjected to an 18-hr chlorination at 15000C, and the volatile thorium tetrachloride released through cracked or very permeable coatings was measured versus chlorination time. Chlorinated batches were also radiographed to detect any thorium that had migrated from the kernel into the coatings. From this work a technique was developed to determine coating permeability. This consists of an 18-hr chlorination of multiple samples without measurement of the heavy metal released. Each batch is then radiographed and the heavy metal diffusion within each particle is examined so it can be determined if a particle batch is permeable, slightly permeable, or gas-tight. (author)

  3. Micro Cooling, Heating, and Power (Micro-CHP) and Bio-Fuel Center, Mississippi State University

    Energy Technology Data Exchange (ETDEWEB)

    Louay Chamra

    2008-09-26

    specifications is observed. Case study data for various micro-CHP system configurations have been discussed and compared. Comparisons are made of the different prime mover/fuel combinations. Also, micro- CHP monthly energy cost results are compared for each system configuration to conventional monthly utility costs for equivalent monthly building power, heating, and cooling requirements.

  4. Dynamic model of a micro-tubular solid oxide fuel cell stack including an integrated cooling system

    Science.gov (United States)

    Hering, Martin; Brouwer, Jacob; Winkler, Wolfgang

    2017-02-01

    A novel dynamic micro-tubular solid oxide fuel cell (MT-SOFC) and stack model including an integrated cooling system is developed using a quasi three-dimensional, spatially resolved, transient thermodynamic, physical and electrochemical model that accounts for the complex geometrical relations between the cells and cooling-tubes. The modeling approach includes a simplified tubular geometry and stack design including an integrated cooling structure, detailed pressure drop and gas property calculations, the electrical and physical constraints of the stack design that determine the current, as well as control strategies for the temperature. Moreover, an advanced heat transfer balance with detailed radiative heat transfer between the cells and the integrated cooling-tubes, convective heat transfer between the gas flows and the surrounding structures and conductive heat transfer between the solid structures inside of the stack, is included. The detailed model can be used as a design basis for the novel MT-SOFC stack assembly including an integrated cooling system, as well as for the development of a dynamic system control strategy. The evaluated best-case design achieves very high electrical efficiency between around 75 and 55% in the entire power density range between 50 and 550 mW /cm2 due to the novel stack design comprising an integrated cooling structure.

  5. A Neutronic Feasibility Study on the Recycling of an Oxide Fuel in Sodium-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok

    2006-06-15

    Neutronic feasibility was implemented for the recycling of a mixed oxide fuel in sodium-cooled fast reactors (SFR) through a thermal/mechanical dry process, which is recognized as one of the most proliferation- resistant recycling processes. In order to assess the applicability of a simple dry process which is not capable of completely removing all the fission products from a spent fuel, sensitivity calculations were performed for the reactor physics parameters with a dependency on the fission product removal rate of the recycled spent fuel. The equilibrium core calculations were performed by the REBUS-3 code for a BN-600 core without blanket fuels and a modified core with an increased fuel volume fraction. The reactor performance parameters such as the transuranic content, breeding ratio, peak linear power, burnup reactivity swing and reactivity coefficients were calculated for an equilibrium core under a fixed fuel management scheme. The results showed that a recycling of the oxide fuel in the SFR is feasible if the fission products are removed by more than 70% through the dry process as far as the material balance is concerned. However the physics analysis also showed that some of the physics design parameters are slightly deteriorated. The results of this study indicate that the recycling characteristics can be improved if the dry process can remove more fission products, and the reactor configuration is further optimized or the spent fuel composition is adjusted.

  6. A Neutronic Feasibility Study on the Recycling of an Oxide Fuel in Sodium-Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Choi, Hang Bok

    2006-06-01

    Neutronic feasibility was implemented for the recycling of a mixed oxide fuel in sodium-cooled fast reactors (SFR) through a thermal/mechanical dry process, which is recognized as one of the most proliferation- resistant recycling processes. In order to assess the applicability of a simple dry process which is not capable of completely removing all the fission products from a spent fuel, sensitivity calculations were performed for the reactor physics parameters with a dependency on the fission product removal rate of the recycled spent fuel. The equilibrium core calculations were performed by the REBUS-3 code for a BN-600 core without blanket fuels and a modified core with an increased fuel volume fraction. The reactor performance parameters such as the transuranic content, breeding ratio, peak linear power, burnup reactivity swing and reactivity coefficients were calculated for an equilibrium core under a fixed fuel management scheme. The results showed that a recycling of the oxide fuel in the SFR is feasible if the fission products are removed by more than 70% through the dry process as far as the material balance is concerned. However the physics analysis also showed that some of the physics design parameters are slightly deteriorated. The results of this study indicate that the recycling characteristics can be improved if the dry process can remove more fission products, and the reactor configuration is further optimized or the spent fuel composition is adjusted

  7. Experimental Study of Using Emulsified Diesel Fuel on the Performance and Pollutants Emitted from Four Stroke Water Cooled Diesel Engine

    Science.gov (United States)

    Sakhrieh, A.; Fouad, R. H.; Yamin, J. A.

    2009-08-01

    A water-cooled, four stroke, four cylinder, direct injection diesel engine was used to study the effect of emulsified diesel fuel on the engine performance and on the main pollutant emissions. Emulsified diesel fuels of 0%, 5%, 10%, 15%, 20%, 25% and 30% water by volume were used. The experiments were conducted in the speed range from 1000 to 3000 rpm. It was found that, in general, using emulsified fuel improves the engine performance and reduces emissions. While the BSFC has a minimum value at 5% water and 2000 rpm, the torque, the BMEP and efficiency are found to have maximum values under these conditions. CO2 was found to increase with engine speed and to decrease with water content. NOx produced from emulsified fuel is significantly less than that produced from pure diesel under the same conditions.

  8. About calculation results of heat transfer in the fuel assembly clusters cooled by water with supercritical parameters

    International Nuclear Information System (INIS)

    Grabezhnaya, V.A.

    2008-01-01

    Paper reviews the numerical investigation into the heat transfer in the supercritical water cooled fuel assemblies on the basis of the various commercial codes. The turbulence available models specified in the codes describe adequately the experimental data in tubes within the range of flow temperatures away from the pseudocritical point, as well as under high mass velocities. There are k-ε type turbulence models that show qualitatively the local acceleration (slowdown) of the heat transfer in tubes, but they fail to describe the mentioned phenomena quantitatively. To determine the effect of grid spacers on the suppression of the heat transfer local slowdown and on the heat transfer acceleration in fuel assemblies and to ensure more accurate calculation of the fuel element cladding maximum temperature one should perform a number of the experiments making use of the fuel assembly models [ru

  9. Development of variable width ribbon heating elements for liquid metal and gas-cooled fast breeder reactor fuel rod simulators

    International Nuclear Information System (INIS)

    McCulloch, R.W.; Lovell, R.T.; Post, D.W.; Snyder, S.D.

    1980-01-01

    Variable width ribbon heating elements have been fabricated which provide a chopped cosine, variable heat flux profile for fuel rod simulators used in test loops by the Breeder Reactor Program Thermal Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor Core Flow Test Loop. Thermal, mechanical, and electrical design considerations result in the derivation of an analytical expression for the ribbon contours. From this, the ribbons are machined and wound on numerically controlled equipment. Postprocessing and inspection results in a wound, variable width ribbon with the precise dimensional, electrical, and mechanical properties needed for use in fuel pin simulators

  10. Disintegration of graphite matrix from the simulative high temperature gas-cooled reactor fuel element by electrochemical method

    International Nuclear Information System (INIS)

    Tian Lifang; Wen Mingfen; Li Linyan; Chen Jing

    2009-01-01

    Electrochemical method with salt as electrolyte has been studied to disintegrate the graphite matrix from the simulative high temperature gas-cooled reactor fuel elements. Ammonium nitrate was experimentally chosen as the appropriate electrolyte. The volume average diameter of disintegrated graphite fragments is about 100 μm and the maximal value is less than 900 μm. After disintegration, the weight of graphite is found to increase by about 20% without the release of a large amount of CO 2 probably owing to the partial oxidation to graphite in electrochemical process. The present work indicates that the improved electrochemical method has the potential to reduce the secondary nuclear waste and is a promising option to disintegrate graphite matrix from high temperature gas-cooled reactor spent fuel elements in the head-end of reprocessing.

  11. Thermal management optimization of an air-cooled hydrogen fuel cell system in an extreme environmental condition

    DEFF Research Database (Denmark)

    Gao, Xin; Olesen, Anders Christian; Kær, Søren Knudsen

    2018-01-01

    An air-cooled proton exchange membrane (PEM) fuel cell system is designed and under manufacture for telecommunication back-up power. To enhance its competence in various environments, the system thermal feature is optimized in this work via simulation based on a computational fluid dynamics (CFD......, the intake airflow magnitude, is also studied for a more uniform airflow and in turn a suppressed temperature disparity inside the system. Following the guidelines drawn by this work on the system design and the operation setting, the air-cooled fuel cell system can be expected with better performances......) model. The model is three-dimensional (3D) and built in the commercial CFD package Fluent (ANSYS Inc.). It makes the full-scale system-level study feasible by only considering the system essences with adequate accuracy. Through the model, the optimization is attained in several aspects. Firstly...

  12. Full-fluence tests of experimental thermosetting fuel rods for the high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Bullock, R.E.

    1981-01-01

    The irradiation performance of injected thermosetting fuel rods is compared to that of standard pitch-temperature gas-cooled reactor requirements. The primary objective of the experiments reported here was to obtain additional irradiation data at higher fluences for resin-based rods with intermediate binder char contents within the 15 to 30 wt% ''window of acceptability'' that had been previously established. 12 refs

  13. Experience from the transportation of irradiated WWER-440 fuel assemblies at the Kozloduy NPP site after a short cooling time

    International Nuclear Information System (INIS)

    Stoyanova, I.; Kamenov, A.; Byrzev, L.; Christoskov, I.

    2003-01-01

    Results from the analysis and experimental verification of the radiation and shielding characteristics of non-standard loading patterns of the VSPOT transport cask used for transportation of irradiated fuel assemblies after a short cooling time (120 180 days) on the site of the Kozloduy NPP are presented. An additional safety criterion related to the introduced modifications to the standard procedure of using the transport cask is formulated and discussed (Authors)

  14. 2D and 3D Modeling Efforts in Fuel Film Cooling of Liquid Rocket Engines (Conference Paper with Briefing Charts)

    Science.gov (United States)

    2017-01-12

    to determine what parameters drive unsteadiness in fuel films, and how these parameters affect wall temperature profiles. Parametric studies performed...temperature profiles. Parametric studies performed in 2D suggest that a Helmholtz resonator exists for simple slot geometries. Frequencies in 3D were...effect on film cooling effectiveness. In general, the heat flux exhibits complex trends and did not scale well with chamber pressure. ∗Aerospace

  15. Long-Term Durability of PBI-Based HT-PEM Fuel Cells: Effect of Operating Parameters

    DEFF Research Database (Denmark)

    Søndergaard, Tonny; Cleemann, Lars Nilausen; Becker, Hans

    2018-01-01

    -term operation. Durability is assessed for 16 identically fabricated membrane electrode assemblies (MEAs), and evaluations are carried out using operating parameters as stressors with gas stoichiometries ranging from 2 to 25, current densities from 200 to 800 mA cm(-2), and temperatures of 160 or 180 degrees C...

  16. Flow-Induced Vibration Measurement of an Inner Cladding Tube in a Simulated Dual-Cooled Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kang Hee; Kim, Hyung Kyu; Yoon, Kyung Ho; Lee, Young Ho; Kim, Jae Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    To create an internal coolant flow passage in a dual cooled fuel rod, an inner cladding tube cannot have intermediate supports enough to relieve its vibration. Thus it can be suffered from a flow-induced vibration (FIV) more severely than an outer cladding tube which will be supported by series of spacer grids. It may cause a fatigue failure at welding joints on the cladding's end plug or fluid elastic instability of long, slender inner cladding due to decrease of a critical flow velocity. This is one of the challenging technical issues when a dual cooled fuel assembly is to be realized into a conventional reactor core To study an actual vibration phenomenon of a dual cooled fuel rod, FIV tests using a small-scale test bundle are being carried out. Measurement results of inner cladding tube of two typically simulated rods are presented. Causes of the differences in the vibration amplitude and response spectrum of the inner cladding tube in terms of intermediate support condition and pellet stacking are discussed.

  17. Design and neutronic investigation of the Nano fluids application to VVER-1000 nuclear reactor with dual cooled annular fuel

    International Nuclear Information System (INIS)

    Ansarifar, G.R.; Ebrahimian, M.

    2016-01-01

    Highlights: • The change in neutronic parameters to the use of nanofluid as coolant is presented. • Nanoparticle deposition on fuel clad is investigated. • Radial and axial local power peaking factors are presented. • ZrO 2 and Al 2 O 3 have the lowest rate of K eff drop off. - Abstract: Nowadays, many efforts have been made to improve the efficiency of nuclear power plants. One of which is use of the dual cooled annular fuel which is an internally and externally cooled annular fuel with many advantages in heat transfer characteristics. Another is the use of nanoparticle/water (nanofluid) as coolant. In this paper, by combining these two methods, the change in neutronic parameters of the VVER-1000 nuclear reactor core with dual cooled annular fuel attributable to the use of nanoparticle/water (nanofluid) as coolant is presented. Optimization of type and volume fraction of nanoparticles in water that affect the safety enhancement of core primary parameters is intended in this study. Reactivity change, radial and axial local power peaking factors (LPPF), and the consequence of nanoparticle deposition on fuel clad are investigated. As a result of changing the effective multiplication factor and PPF calculations for six types of nanoparticles which have been studied extensively for their heat transfer properties including Alumina, Aluminum, Copper oxide, Copper, Titania, and Zirconia with different volume fractions, it can be concluded that at low concentration (0.03 volume fraction), Zirconia and Alumina are the optimum nanoparticles for normal operation. The maximum radial and axial PPF are found to be invariant to the type of nanofluid at low volume fractions. With an increase in nanoparticle deposition thickness on the outer and inner clad, a flux and K eff depression occurred and ZrO 2 and Al 2 O 3 have the lowest rate of drop off.

  18. Treatment and Disposal of the Radioactive Graphite Waste of High-Temperature Gas-Cooled Reactor Spent Fuel

    International Nuclear Information System (INIS)

    Li Junfeng

    2016-01-01

    High-temperature gas-cooled reactors (HTGRs) represent one of the Gen IV reactors in the future market, with efficient generation of energy and the supply of process heat at high temperature utilised in many industrial processes. HTGR development has been carried out within China’s National High Technology Research and Development Program. The first industrial demonstration HTGR of 200 MWe is under construction in Shandong Province China. HTGRs use ceramic-coated fuel particles that are strong and highly resistant to irradiation. Graphite is used as moderator and helium is used as coolant. The fuel particles and the graphite block in which they are imbedded can withstand very high temperature (up to ~1600℃). Graphite waste presents as the fuel element components of HTGR with up to 95% of the whole element beside the graphite blocks in the core. For example, a 200 MWe reactor could discharge about 90,000 fuel elements with 17 tonnes irradiated graphite included each year. The core of the HTGR in China consists of a pebble bed with spherical fuel elements. The UO 2 fuel kernel particles (0.5mm diameter) (triple-coated isotropic fuel particles) are coated by several layers including inner buffer layer with less dense pyrocarbon, dense pyro-carbon, SiC layer and outer layer of dense pyro-carbon, which can prevent the leaking of fission products (Fig. 1). Spherical fuel elements (60mm diameter) consist of a 50mm diameter inner zone and 5mm thick shell of fuel free zone [3]. The inner zone contains about 8300 triple-coated isotropic fuel particles of 0.92mm in diameter dispersed in the graphite matrix

  19. Dynamic Model of the High Temperature Proton Exchange Membrane Fuel Cell Stack Temperature

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Kær, Søren Knudsen

    2009-01-01

    The present work involves the development of a model for predicting the dynamic temperature of a high temperature proton exchange membrane (HTPEM) fuel cell stack. The model is developed to test different thermal control strategies before implementing them in the actual system. The test system co...... elements for start-up, heat conduction through stack insulation, cathode air convection, and heating of the inlet gases in the manifold. Various measurements are presented to validate the model predictions of the stack temperatures....

  20. Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Bartine, D.E.

    1981-01-01

    A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes (1000 and 3000 MW(t)) and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950/sup 0/C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950/sup 0/C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG.

  1. Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kasten, P.R.; Bartine, D.E.

    1981-01-01

    A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes [1000 and 3000 MW(t)] and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950 0 C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950 0 C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG

  2. Aerodynamic and thermal studies of cans of gas cooled fuel elements

    International Nuclear Information System (INIS)

    Gelin, P.

    1964-01-01

    Research on clusters was undertaken at the CEA in 1959, while studies on herring-bone cans were developed at the EDF Laboratory at Chatou and the CEA laboratory at Saclay at the end of 1959. In 1962, a general study on corrugations was begun at the Saclay Laboratory with a view to improving the clusters, and continued later in both laboratories relative to the internal cooling of annular fuel elements. As these studies progressed, trial facilities were extended while experimental methods have improved constantly. At the present time, both laboratories, working in complete collaboration, have powerful means at their disposal. Work on the clusters has been concerned chiefly with pressure losses due to the assembly parts, and with the temperature variations around the elements of the cluster. In this way, we have been able to determine satisfactorily the hot points of the can, the deformations of the rods and the conditions of stability of these deformations. In the case of the herring-bone cans, studies have been directed to the evolution of performances as a function of the geometric parameters on the one hand, and to the special aerodynamic and thermal features caused by the fins and by interruptions of cartridge on the other hand. These studies have led to a very thorough knowledge of the cartridges chosen for the reactors EDF 2 and EDF 3, and now open up very hopeful prospects for future reactors, particularly those fitted with annular elements; among the alternatives suitable for the inner surface of the annular element can, corrugations and longitudinal fins have been fairly extensively tested over a wide range of Reynolds number. (authors) [fr

  3. Status Report on Spent Fuel Pools under Loss-of-Cooling and Loss-of-Coolant Accident Conditions - Final Report

    International Nuclear Information System (INIS)

    Adorni, M.; Esmaili, H.; Grant, W.; Hollands, T.; Hozer, Z.; Jaeckel, B.; Munoz, M.; Nakajima, T.; Rocchi, F.; Strucic, M.; ); Tregoures, N.; Vokac, P.; Ahn, K.I.; Bourgue, L.; Dickson, R.; Douxchamps, P.A.; Herranz, L.E.; Jernkvist, L.O.; Amri, A.; Kissane, M.P.; )

    2015-01-01

    Following the 2011 accident at the Fukushima Daiichi Nuclear Power Station, the Nuclear Energy Agency Committee on the Safety of Nuclear Installations decided to launch several high-priority activities to address certain technical issues. Among other things, it was decided to prepare a status report on spent fuel pools (SFPs) under loss of cooling accident conditions. This activity was proposed jointly by the CSNI Working Group on Analysis and Management of Accidents (WGAMA) and the Working Group on Fuel Safety (WGFS). The main objectives, as defined by these working groups, were to: - Produce a brief summary of the status of SFP accident and mitigation strategies, to better contribute to the post-Fukushima accident decision making process; - Provide a brief assessment of current experimental and analytical knowledge about loss of cooling accidents in SFPs and their associated mitigation strategies; - Briefly describe the strengths and weaknesses of analytical methods used in codes to predict SFP accident evolution and assess the efficiency of different cooling mechanisms for mitigation of such accidents; - Identify and list additional research activities required to address gaps in the understanding of relevant phenomenological processes, to identify where analytical tool deficiencies exist, and to reduce the uncertainties in this understanding. The proposed activity was agreed and approved by CSNI in December 2012, and the first of four meetings of the appointed writing group was held in March 2013. The writing group consisted of members of the WGAMA and the WGFS, representing the European Commission and the following countries: Belgium, Canada, Czech Republic, France, Germany, Hungary, Italy, Japan, Korea, Spain, Sweden, Switzerland and the USA. This report mostly covers the information provided by these countries. The report is organised into 8 Chapters and 4 Appendices: Chapter 1: Introduction; Chapter 2: Spent fuel pools; Chapter 3: Possible accident

  4. Application of Sub-cooled Boiling Model to Thermal-hydraulic Analysis Inside a CANDU-6 Fuel Channel

    International Nuclear Information System (INIS)

    Kim, Man Woong; Lee, Sang Kyu; Kim, Hyun Koon; Yoo, Kun Joong; Kang, Hyoung Chul; Yoo, Seong Yeon

    2007-01-01

    Forced convection nucleate boiling is encountered in heat exchangers during normal and non-nominal modes of operation in pressurized water or boiling water reactors (PWRs or BWRs). If the wall temperature of the piping is higher than the saturation temperature of the nearby liquid, nucleate boiling occurs. In this regime, bubbles are formed at the wall. Their growth is promoted by the wall superheat (the difference between the wall and saturation temperatures), and they depart from the wall as a result of gravitational and liquid inertia forces. If the bulk liquid is subcooled, condensation at the bubble-liquid interface takes place and the bubble may collapse. This convection nucleate boiling is called as a sub-cooled nucleate boiling. As for the fuel channel of a CANDU 6 reactor, forced convection nucleate boiling models for flows along fuel elements enclosed inside typical CANDU-6 fuel channel has encountered difficulties due to the modeling of local effects along the horizontal channel. Therefore, the subcooled nucleate boiling has been modeled through temperature driven boiling heat and mass transfer, using a model developed at Rensselaer Polytechnic Institute. The objectives of this study are: (i) to investigate a proposed sub-cooled boiling model developed at Rensselaer Polytechnic Institute and (ii) to apply against a experiment and (iii) to predict local distributions of flow fields for the actual fuel channel geometries of CANDU-6 reactors. The numerical implementation is conducted using by the FLUENT 6.2 CFD computer code

  5. Effects of Cooling Fluid Flow Rate on the Critical Heat Flux and Flow Stability in the Plate Fuel Type 2 MW TRIGA Reactor

    OpenAIRE

    H. P. Rahardjo; V. I. Sri Wardhani

    2017-01-01

    The conversion program of the 2 MW TRIGA reactor in Bandung consisted of the replacement of cylindrical fuel (produced by General Atomic) with plate fuel (produced by BATAN). The replacement led into the change of core cooling process from upward natural convection type to downward forced convection type, and resulted in different thermohydraulic safety criteria, such as critical heat flux (CHF) limit, boiling limit, and cooling fluid flow stability. In this paper, a thermohydraulic safety an...

  6. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    Science.gov (United States)

    Ilham, Muhammad; Su'ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  7. Experience from transportation of irradiated WWER-440 fuel assemblies at Kozloduy NPP site after a short cooling time

    International Nuclear Information System (INIS)

    Stoyanova, I.; Kamenov, A.; Byrzev, L.; Christoskov, I.

    2003-01-01

    The presented results from the computation and analysis of the radiation characteristics of the irradiated fuel assemblies by the date of their transportation according to the selected loading patterns of the VSPOT cask and following the modified technology of transportation, i.e. without replacement of the pool solution by pure condensate, as well as the corresponding experimental results, confirm the applicability of the newly introduced safety criterion for the selection of a loading pattern of the cask with irradiated fuel assemblies after a short cooling time. The comparison between measured and computed surface dose rates shows that during the procedure of transfer of irradiated fuel assemblies from the pools of Units 1 and 2 to the pools of Kozloduy NPP Units 3 and 4 all safety limits, incl. the radiation protection requirements, were met

  8. Neutronic and Isotopic Simulation of a Thorium-TRU's fuel Closed Cycle in a Lead Cooled ADS

    International Nuclear Information System (INIS)

    Garcia-Sanz, J. M.; Embid, M.; Fernandez, R.; Gonzalez, E. M.; Perez-Parra, A.

    2000-01-01

    The FACET group at CIEMAT is studying the properties and potentialities of several lead-cooled ADS designs for actinide and fission product transmutation. The main characteristics of these systems are the use of lead as primary coolant and moderator and fuels made by transuranics inside a thorium oxide matrix. The strategy assumed in this simulation implies that every discharge of the ADS will be reprocessed and would produce four waste streams: fission and activation products, remaining ''232 Th, produced ''233 U and remaining TRU's. The ''233 U is separated for other purposes; the remaining TRU are recovered altogether and mixed with the adequate amount of ''232 Th and fresh TRUs coming from LWR spent fuel. The simulations performed in this study have been focused primarily in the evolution of the fuel isotopic composition during and after each ADS burn-up cycle. (Author) 10 refs

  9. Thermodynamic Data to Model the Interaction Between Coolant and Fuel in Gen IV Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Dinsdale, Alan; Gisby, John; Davies, Hugh; Konings, Rudy; Benes, Ondrej

    2013-06-01

    Understanding the behaviour of nuclear fuels in various environments is vital to the design and safe operation of nuclear reactors. While this is true if the reactor is operating within its design specification, it is even more so if accidents occur and the fuel is exposed to unexpected temperatures, pressures or chemical environments. It is clearly hazardous and costly to explore all such scenarios experimentally and therefore it is necessary to undertake modelling where possible using well-grounded theoretical approaches. This paper will show examples of where calculations of chemical and phase equilibria have been applied successfully to the long term storage of nuclear waste, phase formation during core meltdown and prediction of fission product release into the atmosphere. It will also highlight the development of thermodynamic data carried out during the European Metrology Research Project Metrofission required to model the potential interaction between the coolant, nuclear fuel, containment materials and atmosphere of a sodium cooled fast reactor. (authors)

  10. Process for changing fuel elements of a water-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Fleischmann, R.; Rau, P.

    1986-01-01

    In order to change fuel elements, a water-filled duct can be installed between the rector pressure vessel and a space for accommodating the fuel elements. The fuel elements are transported there under water by a fuelling machine. The duct is installed as watertight connection closed on all sides between the reactor pressure vessel and a fuel element transport container brought close to it. The fuelling machine works in this duct. (orig./HP) [de

  11. Primary energy savings in desiccant and evaporative cooling-assisted 100% outdoor air system combined with a fuel cell

    International Nuclear Information System (INIS)

    Kim, Min-Hwi; Dong, Hae-Won; Park, Joon-Young; Jeong, Jae-Weon

    2016-01-01

    Highlights: • A LD-IDECOAS integrated with a PEMFC was proposed. • A pilot system was installed and tested during cooling operation. • The proposed system powered by the PEMFC saved 21% of the primary energy consumption during cooling. - Abstract: The main purpose of this study involved investigating the primary energy saving potential of a liquid desiccant and evaporative cooling-assisted 100% outdoor air system (LD-IDECOAS) integrated with a proton exchange membrane fuel cell (PEMFC). During the cooling season, the heat produced by the PEMFC was used to regenerate a weak desiccant solution, and the electricity generated was used to operate the LD-IDECOAS. A pilot LD-IDECOAS powered by a PEMFC was installed and operated in an office space to experimentally verify the annual operating energy savings of the proposed system. The findings indicated that the heat reclaimed from the PEMFC saved 42% of the desiccant solution regenerating energy when compared to that in the case of a conventional gas-fired water heater. The results also suggested that the LD-IDECOAS combined with a PEMFC consumed 21% less primary energy when compared with that of a system powered by grid electricity and a conventional gas-fired water heater.

  12. Modeling and Application of Pneumatic Conveying for Spherical Fuel Element in Pebble-Bed Modular High-Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Zhou Shuyong; Wang Junsan; Wang Yuding; Cai Ruizhong; Zhang Xuan; Cao Jianting

    2014-01-01

    The fuel handling system is an important system for on-load refueling in pebble-bed modular high-temperature gas-cooled reactor. A dynamic model of pneumatic conveying for spherical fuel element in fuel handling system was established to describe the pneumatically conveying process. The motion characteristics of fuel elements in pipeline and the effect of fuel elements on gas velocity were studied using the model. The results show that the theoretical analyses are consistent with the experimental. The research has been used in developing full scope simulator for pebble-bed modular high-temperature gas-cooled reactor, also provides references for the design and optimization of the fuel handling system. (author)

  13. Stoichiometric effects on performance of high-temperature gas-cooled reactor fuels from the U--C--O system

    International Nuclear Information System (INIS)

    Homan, F.J.; Lindemer, T.B.; Long, E.L. Jr.; Tiegs, T.N.; Beatty, R.L.

    1977-01-01

    Two fuel failure mechanisms were identified for coated particle fuels that are directly related to fuel kernel stoichiometry. These mechanisms are thermal migration of the kernel through the coating layers and chemical interaction between rare-earth fission products and the silicon carbide (SiC) layer leading to failure of the SiC layer. Thermal migration appears to be most severe for oxide fuels, while chemical interaction is most severe with carbide systems. Thermodynamic calculations indicated that oxide-carbide fuel kernels may permit a stoichiometry that reduces both problems to manageable levels for currently planned high-temperature gas-cooled reactors. Such stoichiometry adjustment is possible over the complete spectrum from UO 2 to UC 2 for the present recycle fuel, a weak acid resin (WAR)-derived fissile kernel. Thermodynamic calculations indicate that WAR kernels containing less than 15 percent UC 2 (greater than 85 percent UO 2 ) will develop excessive CO overpressures within the particle during irradiation. In 100 percent UO 2 particles, thermal migration and oxidation of the SiC layer were observed after irradiation. The calculations also indicate that WAR kernels containing greater than 70 percent UC 2 (less than 30 percent UC 2 ) contain insufficient oxygen to oxidize the rare-earth fission products formed in fuel operated to the maximum burnup levels of 75 percent fissions per initial metal atom (75 percent FIMA). Instead, the rare earths are present in part or completely as dicarbides. As such, they were observed to segregate from the kernel and collect at the SiC interface on the cold side of the particle, react with the SiC, and eventually fail this coating

  14. Upper limits to americium concentration in large sized sodium-cooled fast reactors loaded with metallic fuel

    International Nuclear Information System (INIS)

    Zhang, Youpeng; Wallenius, Janne

    2014-01-01

    Highlights: • The americium transmutation capability of Integral Fast Reactor was investigated. • The impact from americium introduction was parameterized by applying SERPENT Monte Carlo calculations. • Higher americium content in metallic fuel leads to a power penalty, preserving consistent safety margins. - Abstract: Transient analysis of a large sized sodium-cooled reactor loaded with metallic fuel modified by different fractions of americium have been performed. Unprotected loss-of-offsite power, unprotected loss-of-flow and unprotected transient-over-power accidents were simulated with the SAS4A/SASSYS code based on the geometrical model of an IFR with power rating of 2500 MW th , using safety parameters obtained with the SERPENT Monte Carlo code. The Ti-modified austenitic D9 steel, having higher creep rupture strength, was considered as the cladding and structural material apart from the ferritic/martensitic HT9 steel. For the reference case of U–12Pu–1Am–10Zr fuel at EOEC, the margin to fuel melt during a design basis condition UTOP is about 50 K for a maximum linear rating of 30 kW/m. In order to maintain a margin of 50 K to fuel failure, the linear power rating has to be reduced by ∼3% and 6% for 2 wt.% and 3 wt.% Am introduction into the fuel respectively. Hence, an Am concentration of 2–3 wt.% in the fuel would lead to a power penalty of 3–6%, permitting a consumption rate of 3.0–5.1 kg Am/TW h th . This consumption rate is significantly higher than the one previously obtained for oxide fuelled SFRs

  15. Heat transfer on HLM cooled wire-spaced fuel pin bundle simulator in the NACIE-UP facility

    Energy Technology Data Exchange (ETDEWEB)

    Di Piazza, Ivan, E-mail: ivan.dipiazza@enea.it [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone, Camugnano (Italy); Angelucci, Morena; Marinari, Ranieri [University of Pisa, Dipartimento di Ingegneria Civile e Industriale, Pisa (Italy); Tarantino, Mariano [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone, Camugnano (Italy); Forgione, Nicola [University of Pisa, Dipartimento di Ingegneria Civile e Industriale, Pisa (Italy)

    2016-04-15

    Highlights: • Experiments with a wire-wrapped 19-pin fuel bundle cooled by LBE. • Wall and bulk temperature measurements at three axial positions. • Heat transfer and error analysis in the range of low mass flow rates and Péclet number. • Comparison of local and section-averaged Nusselt number with correlations. - Abstract: The NACIE-UP experimental facility at the ENEA Brasimone Research Centre (Italy) allowed to evaluate the heat transfer coefficient of a wire-spaced fuel bundle cooled by lead-bismuth eutectic (LBE). Lead or lead-bismuth eutectic are very attractive as coolants for the GEN-IV fast reactors due to the good thermo-physical properties and the capability to fulfil the GEN-IV goals. Nevertheless, few experimental data on heat transfer with heavy liquid metals (HLM) are available in literature. Furthermore, just a few data can be identified on the specific topic of wire-spaced fuel bundle cooled by HLM. Additional analysis on thermo-fluid dynamic behaviour of the HLM inside the subchannels of a rod bundle is necessary to support the design and safety assessment of GEN. IV/ADS reactors. In this context, a wire-spaced 19-pin fuel bundle was installed inside the NACIE-UP facility. The pin bundle is equipped with 67 thermocouples to monitor temperatures and analyse the heat transfer behaviour in different sub-channels and axial positions. The experimental campaign was part of the SEARCH FP7 EU project to support the development of the MYRRHA irradiation facility (SCK-CEN). Natural and mixed circulation flow regimes were investigated, with subchannel Reynolds number in the range Re = 1000–10,000 and heat flux in the range q″ = 50–500 kW/m{sup 2}. Local Nusselt numbers were calculated for five sub-channels in different ranks at three axial positions. Section-averaged Nusselt number was also defined and calculated. Local Nusselt data showed good consistency with some of the correlation existing in literature for heat transfer in liquid metals

  16. Heat transfer on HLM cooled wire-spaced fuel pin bundle simulator in the NACIE-UP facility

    International Nuclear Information System (INIS)

    Di Piazza, Ivan; Angelucci, Morena; Marinari, Ranieri; Tarantino, Mariano; Forgione, Nicola

    2016-01-01

    Highlights: • Experiments with a wire-wrapped 19-pin fuel bundle cooled by LBE. • Wall and bulk temperature measurements at three axial positions. • Heat transfer and error analysis in the range of low mass flow rates and Péclet number. • Comparison of local and section-averaged Nusselt number with correlations. - Abstract: The NACIE-UP experimental facility at the ENEA Brasimone Research Centre (Italy) allowed to evaluate the heat transfer coefficient of a wire-spaced fuel bundle cooled by lead-bismuth eutectic (LBE). Lead or lead-bismuth eutectic are very attractive as coolants for the GEN-IV fast reactors due to the good thermo-physical properties and the capability to fulfil the GEN-IV goals. Nevertheless, few experimental data on heat transfer with heavy liquid metals (HLM) are available in literature. Furthermore, just a few data can be identified on the specific topic of wire-spaced fuel bundle cooled by HLM. Additional analysis on thermo-fluid dynamic behaviour of the HLM inside the subchannels of a rod bundle is necessary to support the design and safety assessment of GEN. IV/ADS reactors. In this context, a wire-spaced 19-pin fuel bundle was installed inside the NACIE-UP facility. The pin bundle is equipped with 67 thermocouples to monitor temperatures and analyse the heat transfer behaviour in different sub-channels and axial positions. The experimental campaign was part of the SEARCH FP7 EU project to support the development of the MYRRHA irradiation facility (SCK-CEN). Natural and mixed circulation flow regimes were investigated, with subchannel Reynolds number in the range Re = 1000–10,000 and heat flux in the range q″ = 50–500 kW/m"2. Local Nusselt numbers were calculated for five sub-channels in different ranks at three axial positions. Section-averaged Nusselt number was also defined and calculated. Local Nusselt data showed good consistency with some of the correlation existing in literature for heat transfer in liquid metals for

  17. A Multi-function Cask for At-Reactor Storage of Short-Cooled Spent Fuel, Transport, and Disposal

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    2004-01-01

    The spent nuclear fuel (SNF) system in the United States was designed with the assumptions that SNF would be stored for several years in an at-reactor pool and then transported to reprocessing plants for recovery of fissile materials, that security would not be a major issue, and that the SNF burnups would be low. The system has evolved into a once-through fuel cycle with high-burnup SNF, long-term storage at the reactor sites, and major requirements for safeguards and security. An alternative system is proposed to better meet these current requirements. The SNF is placed in multi-function casks with the casks used for at-reactor storage, transport, and repository disposal. The cask is the handling package, provides radiation shielding, and protects the SNF against accidents and assault. SNF assemblies are handled only once to minimize accident risks, maximize security and safeguards by minimizing access to SNF, and reduce costs. To maximize physical protection, the cask body is constructed of a cermet (oxide particles embedded in steel, the same class of materials used in tank armor) and contains no cooling channels or other penetrations that allow access to the SNF. To minimize pool storage of SNF, the cask is designed to accept short-cooled SNF. To maximize the capability of the cask to reject decay heat and to limit SNF temperatures from short-cooled SNF, the cask uses (1) natural circulation of inert gas mixtures inside the cask to transfer heat from the SNF to the cask body and (2) an overpack with external natural-circulation, liquid-cooled fins to transfer heat from the cask body to the atmosphere. This approach utilizes the entire cask body area for heat transfer to maximize heat removal rates-without any penetrations through the cask body that would reduce the physical protection capabilities of the cask body. After the SNF has cooled, the cooling overpack is removed. At the repository, the cask is placed in a corrosion-resistant overpack before disposal

  18. Proposed master-slave and automated remote handling system for high-temperature gas-cooled reactor fuel refabrication

    International Nuclear Information System (INIS)

    Grundmann, J.G.

    1974-01-01

    The Oak Ridge National Laboratory's Thorium-Uranium Recycle Facility (TURF) will be used to develop High-Temperature Gas-Cooled Reactor (HTGR) fuel recycle technology which can be applied to future HTGR commercial fuel recycling plants. To achieve recycle capabilities it is necessary to develop an effective material handling system to remotely transport equipment and materials and to perform maintenance tasks within a hot cell facility. The TURF facility includes hot cells which contain remote material handling equipment. To extend the capabilities of this equipment, the development of a master-slave manipulator and a 3D-TV system is necessary. Additional work entails the development of computer controls to provide: automatic execution of tasks, automatic traverse of material handling equipment, automatic 3D-TV camera sighting, and computer monitoring of in-cell equipment positions to prevent accidental collisions. A prototype system which will be used in the development of the above capabilities is presented. (U.S.)

  19. Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

    2007-09-01

    The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean

  20. Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

    2007-09-15

    The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean.

  1. An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

    International Nuclear Information System (INIS)

    Santos, Adimir dos; Nascimento, Jamil Alves do

    2002-01-01

    An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 %Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pinch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

  2. An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

    International Nuclear Information System (INIS)

    Santos, A. dos; Nascimento, J.A. do

    2002-01-01

    An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 % Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pitch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

  3. Multi-criteria methodology to design a sodium-cooled carbide-fueled Gen-IV reactor

    International Nuclear Information System (INIS)

    Stauff, N.

    2011-01-01

    Compared with earlier plant designs (Phenix, Super-Phenix, EFR), Gen IV Sodium-cooled Fast Reactor requires improved economics while meeting safety and non-proliferation criteria. Mixed Oxide (U-Pu)O 2 fuels are considered as the reference fuels due to their important and satisfactory feedback experience. However, innovative carbide (U-Pu)C fuels can be considered as serious competitors for a prospective SFR fleet since carbide-fueled SFRs can offer another type of optimization which might overtake on some aspects the oxide fuel technology. The goal of this thesis is to reveal the potentials of carbide by designing an optimum carbide-fueled SFR with competitive features and a naturally safe behavior during transients. For a French nuclear fleet, a 1500 MW(e) break-even core is considered. To do so, a multi-physic approach was developed taking into account neutronics, fuel thermo-mechanics and thermal-hydraulic at a pre-design stage. Simplified modeling with the calculation of global neutronic feedback coefficients and a quasi-static evaluation was developed to estimate the behavior of a core during overpower transients, loss of flow and/or loss of heat removal transients. The breakthrough of this approach is to provide the designer with an overall view of the iterative process, emphasizing the well-suited innovations and the most efficient directions that can improve the SFR design project.This methodology was used to design a core that benefits from the favorable features of carbide fuels. The core developed is a large carbide-fueled SFR with high power density, low fissile inventory, break-even capability and forgiving behaviors during the un-scrammed transients studied that should prevent using expensive mitigate systems. However, the core-peak burnup is unlikely to significantly exceed 100 MWd/kg because of the large swelling of the carbide fuel leading to quick pellet-clad mechanical interaction and the low creep capacity of carbide. Moderate linear power fuel

  4. Irradiation behaviour of advanced fuel elements for the helium-cooled high temperature reactor (HTR)

    International Nuclear Information System (INIS)

    Nickel, H.

    1990-05-01

    The design of modern HTRs is based on high quality fuel. A research and development programme has demonstrated the satisfactory performance in fuel manufacturing, irradiation testing and accident condition testing of irradiated fuel elements. This report describes the fuel particles with their low-enriched UO 2 kernels and TRISO coating, i.e. a sequence of pyrocarbon, silicon carbide, and pyrocarbon coating layers, as well as the spherical fuel element. Testing was performed in a generic programme satisfying the requirements of both the HTR-MODUL and the HTR 500. With a coating failure fraction less than 2x10 -5 at the 95% confidence level, the results of the irradiation experiments surpassed the design targets. Maximum accident temperatures in small, modular HTRs remain below 1600deg C, even in the case of unrestricted core heatup after depressurization. Here, it was demonstrated that modern TRISO fuels retain all safety-relevant fission products and that the fuel does not suffer irreversible changes. Isothermal heating tests have been extended to 1800deg C to show performance margins. Ramp tests to 2500deg C demonstrate the limits of present fuel materials. A long-term programm is planned to improve the statistical significance of presently available results and to narrow remaining uncertainty limits. (orig.) [de

  5. Numerical prediction on turbulent heat transfer of a spacer ribbed fuel rod for high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Takase, Kazuyuki

    1994-11-01

    The turbulent heat transfer of a fuel rod with three-dimensional trapezoidal spacer ribs for high temperature gas-cooled reactors was analyzed numerically using the k-ε turbulence model, and investigated experimentally using a simulated fuel rod under the helium gas condition of a maximum outlet temperature of 1000degC and pressure of 4MPa. From the experimental results, it found that the turbulent heat transfer coefficients of the fuel rod were 18 to 80% higher than those of a concentric smooth annulus at a region of Reynolds number exceeding 2000. On the other hand, the predicted average Nusselt number of the fuel rod agreed well with the heat transfer correlation obtained from the experimental data within a relative error of 10% with Reynolds number of more than 5000. It was verified that the numerical analysis results had sufficient accuracy. Furthermore, the numerical prediction could clarify quantitatively the effects of the heat transfer augmentation by the spacer rib and the axial velocity increase due to a reduction in the annular channel cross-section. (author)

  6. Acceleration Test Method for Failure Prediction of the End Cap Contact Region of Sodium Cooled Fast Reactor Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung-Kyu; Lee, Young-Ho; Lee, Hyun-Seung; Lee, Kang-Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-05-15

    This paper reports the results of an acceleration test to predict the contact-induced failure that could occur at the cylinder-to-hole joint for the fuel rod of a sodium-cooled fast reactor (SFR). To incorporate the fuel life of the SFR currently under development at KAERI (around 35,000 h), the acceleration test method of reliability engineering was adopted in this work. A finite element method was used to evaluate the flow-induced vibration frequency and amplitude for the test parameter values. Five specimens were tested. The failure criterion during the life of the SFR fuel was applied. The S-N curve of the HT-9, the material of concern, was used to obtain the acceleration factor. As a result, a test time of 16.5 h was obtained for each specimen. It was concluded that the B{sub 0.004} life would be guaranteed for the SFR fuel rods with 99% confidence if no failure was observed at any of the contact surfaces of the five specimens.

  7. Metrological certification of systems to monitor the seal integrity of fuel-element cladding based on exposed fuel in sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Eliseev, A.V.; Filonov, V.S.; Ushakov, V.M.; Belov, S.P.; Pedyash, B.V.; Zemtsev, B.V.; Skorikov, N.V.

    1992-01-01

    In sodium-cooled fast reactors, the clad monitoring system for seal integrity of the fuel element cladding is practically the only source of operator information on the serviceability of fuel elements in the core. The monitoring system can be used as the basis for critical decisions whether the reactor must be shut down of whether operation can continue, but only if the meterologically provided measurements are reliable. This article describes a method developed for certifying working rods on the basis of the domestic standard. The method includes a combined irradiation of the sample and the rod to be certified in an arbitrary field of a plutonium-beryllium neutron source with an output rate greater than 10 8 sec -1 , which is mounted in a paraffin moderator. The positive results of the metrological certification of the system to monitor cladding seal integrity leads the authors to recommend this method for other current and planned sodium-cooled fast reactors. 6 refs., 2 tabs

  8. Advanced fuel pellet materials and designs for water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2004-10-01

    This meeting was the second IAEA meeting on this subject. The first was held in 1996 in Tokyo, Japan. They are all part of a cooperative effort through the Technical Working Group on Water Reactor Fuel Performance and Technology (TWGFPT) of IAEA, with a series of three further meetings organized by CEA, France and co-sponsored by the IAEA and OECD/NEA. In the seven years since the first meeting took place, the demands on fuel duties have increased, with higher burnup, longer fuel cycles and higher temperatures. This places additional demands on fuel performance to comply with safety requirements. Criteria relative to fuel components, i.e. pellets and fuel rod column, require limiting of fission gas release and pellet-cladding interaction (PCI). This means that fuel components should maintain the composite of rather contradictory properties from the beginning until the end of its in-pile operation. Fabrication and design tools are available to influence, and to some extent, to ensure desirable in-pile fuel properties. Discussion of these tools was one of the objectives of the meeting. The second objective was the analysis of fuel characteristics at high burnup and the third and last objective was the discussion of specific feature of MOX and urania gadolinia fuels. Sixty specialists in the field of fuel fabrication technology attended the meeting from 18 countries. Twenty-five papers were presented in five sessions covering all relevant topics from the practices and modelling of fuel fabrication technology to its optimization. Eight papers were presented in session 'Optimization of fuel fabrication technology' which all were devoted to fuel fabrication technology. They mostly treated methods for optimizing fuel manufacturing processes, but gave also a good overview on nuclear fabrication needs and capabilities in different countries. During Session 'UO 2 , MOX and UO 2 -Gd 2 O 3 pellets with additives', six papers were presented in this session, which dealt mainly

  9. Fossil fuel and biomass burning effect on climate - Heating or cooling?

    Science.gov (United States)

    Kaufman, Yoram J.; Fraser, Robert S.; Mahoney, Robert L.

    1991-01-01

    The basic theory of the effect of pollution on cloud microphysics and its global implications is applied to compare the relative effect of a small increase in the consumption rate of oil, coal, or biomass burning on cooling and heating of the atmosphere. The characteristics of and evidence for the SO2 induced cooling effect are reviewed. This perturbation analysis approach permits linearization, therefore simplifying the analysis and reducing the number of uncertain parameters. For biomass burning the analysis is restricted to burning associated with deforestation. Predictions of the effect of an increase in oil or coal burning show that within the present conditions the cooling effect from oil and coal burning may range from 0.4 to 8 times the heating effect.

  10. Thermal-hydraulic study of the LBE-cooled fuel assembly in the MYRRHA reactor: Experiments and simulations

    Energy Technology Data Exchange (ETDEWEB)

    Pacio, J., E-mail: Julio.pacio@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Wetzel, T. [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Doolaard, H.; Roelofs, F. [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 ZG Petten (Netherlands); Van Tichelen, K. [Belgian Nuclear Reseach Center (SCK-CEN), Boeretang 200, Mol (Belgium)

    2017-02-15

    Heavy liquid metals (HLMs), such as lead-bismuth eutectic (LBE) and pure lead are prominent candidate coolants for many advanced systems based on fast neutrons. In particular, LBE is used in the first-of-its-kind MYRRHA fast reactor, to be built in Mol (Belgium), which can be operated either in critical mode or as a sub-critical accelerator-driven system. With a strong focus on safety, key thermal-hydraulic aspects of these systems, such as the proper cooling of fuel assemblies, must be assessed. Considering the complex geometry and low Prandtl number of LBE (Pr ∼ 0.025), this flow scenario is challenging for the models used in Computational Fluid Dynamics (CFD), e.g. for relating the turbulent transport of momentum and heat. Thus, reliable experimental data for the relevant scenario are needed for validation. In this general context, this topic is studied both experimentally and numerically in the framework of the European FP7 project SEARCH (2011–2015). An experimental campaign, including a 19-rod bundle with wire spacers, cooled by LBE is undertaken at KIT. With prototypical geometry and operating conditions, it is intended to evaluate the validity of current empirical correlations for the MYRRHA conditions and, at the same time, to provide validation data for the CFD simulations performed at NRG. The results of one benchmarking case are presented in this work. Moreover, this validated approach is then used for simulating a complete MYRRHA fuel assembly (127 rods).

  11. Exergy analysis of an integrated solid oxide fuel cell and organic Rankine cycle for cooling, heating and power production

    Science.gov (United States)

    Al-Sulaiman, Fahad A.; Dincer, Ibrahim; Hamdullahpur, Feridun

    The study examines a novel system that combined a solid oxide fuel cell (SOFC) and an organic Rankine cycle (ORC) for cooling, heating and power production (trigeneration) through exergy analysis. The system consists of an SOFC, an ORC, a heat exchanger and a single-effect absorption chiller. The system is modeled to produce a net electricity of around 500 kW. The study reveals that there is 3-25% gain on exergy efficiency when trigeneration is used compared with the power cycle only. Also, the study shows that as the current density of the SOFC increases, the exergy efficiencies of power cycle, cooling cogeneration, heating cogeneration and trigeneration decreases. In addition, it was shown that the effect of changing the turbine inlet pressure and ORC pump inlet temperature are insignificant on the exergy efficiencies of the power cycle, cooling cogeneration, heating cogeneration and trigeneration. Also, the study reveals that the significant sources of exergy destruction are the ORC evaporator, air heat exchanger at the SOFC inlet and heating process heat exchanger.

  12. High Temperature PEM Fuel Cells and Organic Fuels

    DEFF Research Database (Denmark)

    Vassiliev, Anton

    of the products. The observation of internal reforming was indirectly confirmed by electrochemical impedance spectroscopy, where the best fits were obtained when a Gerischer element describing preceding chemical reaction and diffusion was included in the equivalent circuit of a methanol/air operated cell...... evaporated liquid stream supply to either of the electrodes. A large number of MEAs with different component compositions have been prepared and tested in different conditions using the constructed setups to obtain a basic understanding of the nature of direct DME HT-PEM FC, to map the processes occurring...... inside the cells and to determine the lifetime. Additionally, comparison was made with methanol as fuel, which is the main competitor to DME in direct oxidation of organic fuels in fuel cells. For the reference, measurements have also been done with conventional hydrogen/air operation. All...

  13. An experimental study on the cathode humidification and evaporative cooling of polymer electrolyte membrane fuel cells using direct water injection method at high current densities

    International Nuclear Information System (INIS)

    Hwang, Seong Hoon; Kim, Min Soo

    2016-01-01

    Highlights: • Proposal of a cathode humidification and evaporative cooling system for PEM fuel cells. • An external-mixing air-assist atomizer is used to produce a very fine water spray. • The system is effective in both cathode humidification and stack cooling. • Increased water flow rate improves stack performance and evaporative cooling capacity. • At a given water flow rate, lower stack temperatures cause greater humidification effect. - Abstract: Humidification and cooling are critical issues in enhancing the efficiency and durability of polymer electrolyte membrane fuel cells (PEMFCs). However, existing humidifiers and cooling systems have the disadvantage that they must be quite large to achieve adequate PEMFC performance. In this study, to eliminate the need for a bulky humidifier and to lighten the cooling load of PEMFCs, a cathode humidification and evaporative cooling system using an external-mixing air-assist atomizer was developed and its performance was investigated. The atomization performance of the nozzle was analyzed experimentally under various operating conditions with minimal changes in the system design. Experiments with a five-cell PEMFC stack with an active area of 250 cm"2 were carried out to analyze the effects of various parameters (such as the operating temperature, current density, and water injection flow rate) on the evaporation of injected water for humidification and cooling performances. The experimental results demonstrate that the direct water injection method proposed in this study is quite effective in cathode humidification and stack cooling in PEM fuel cells at high current densities. The stack performance was improved by humidification effect and the coolant temperature at the stack outlet decreased by evaporative cooling effect.

  14. Nondestructive determination of burnup and fissile isotope balance in spent fuel assemblies of water cooled reactors

    International Nuclear Information System (INIS)

    Pinel, J.

    1983-03-01

    Two non-destructive methods for measuring fuel assemblies in storage pools have been developed: a gamma fuel scanning method, using the 134 Cs - 137 Cs and 144 Ce gamma rays, and the measurement of the neutron flux emitted by the fuel assembly. For interpreting the measurement, we have used calculated correlations to establish a connection between the measured phenomena and the parameters to be determined. A measurement campaign involving 58 assemblies from the C.N.A. reactor was conducted in the reprocessing plant of LA HAGUE. The results obtained show that the objectives can be achevied within an industrial environment [fr

  15. The behaviour of water-cooled reactor fuel rods in steady state and transient conditions

    International Nuclear Information System (INIS)

    Strupczewski, A.; Marks, P.

    1997-01-01

    In this report, the results of temperature field and filling gas pressure calculations by means of contemporary calculational models for a WWER-440 and WWER-1000 type fuel rod at low and high burnup operating under steady-state conditions are presented. A review of in-core temperature and pressure measurements for various types of LWR fuel is also included. Basing on calculational and collected measured data, the behaviour of fuel cladding during large and small break LOCA, is estimated with special emphasis on their oxidation and failure resistance. (author)

  16. Industrial space heating and cooling from stored spent nuclear power plant fuel

    International Nuclear Information System (INIS)

    Shaver, B.O.; Doman, J.W.

    1980-01-01

    Projections by the Department of Energy indicate that some 5800 metric tons of spent fuel from nuclear power reactors are now in storage and that some 33000 metric tons are expected to be in storage in 1990. The bulk of the spent fuel is currently stored in water-filled basins at the reactor sites from which the material was discharged. The thermal energy in the fuel is dissipated to atmospheres via a pumped water-to-air heat exchanger system. This paper describes a feasibility study of potential methods for the use of the heat. Also, potential applications of heat recovery systems at larger AFR storage facilities were investigated

  17. Phytoplankton in the cooling pond of a nuclear fuel plant. II. Spectral analysis

    International Nuclear Information System (INIS)

    Tokarskaya, Z.B.; Smagin, A.I.; Ryzhkov, E.G.; Nikitina, L.V.

    1995-01-01

    This study continues investigations into the development dynamics of phytoplankton and hydrochemical and meteorological factors over a periods of 26 years in the cooling pond of the Mayak Production Association in the Kyzyl-Trash Lake. The aim is to evaluate the long-term oscillations in phytoplankton owing to changes in hydrochemical and meteorological factors. 6 refs., 2 figs., 1 tab

  18. Gas-cooled nuclear reactor with a filling of spherical fuel elements

    International Nuclear Information System (INIS)

    Hantke, H.J.

    1978-01-01

    In order to protect the reflector blanket of a pebble bed reactor against radiation damage a filling of graphite spheres is arranged between blanket and fuel elements, having got a smaller diameter than fuel spheres. Before reaching unduely high irradiation values caused by fast neutrons these graphite spheres are removed from the core, together with the usual discharge of spheres, and replaced by new spheres. (TK) [de

  19. Principles, design and fuel performance characteristics of gas cooled thermal reactors

    International Nuclear Information System (INIS)

    Boocock, P.M.; Eaton, J.R.P.

    1989-01-01

    Reactor output and availability are closely related to fuel design and performance and the SSEB, in collaboration with the Central Electricity Generating Board have followed a policy of continuous analysis and improvement. The position reached is set out and some views on further improvements, are given. The strategy of increasing fuel burn-up on Hunterston A power station has brought significant dividends in the form of major benefits in fuel cycle cost and station availability. Significant improvements in output and availability at Hunterston B have resulted from increasing the fuel cycle burn-up, from 18 GWd/t U to 21 GWd/t U and introducing on-load refuelling. Additional benefits are soon to be obtained by further extending the burn-up to 24 GWd/t U. Further reduction of typically Pound 2-7 million/year in fuel cycle costs over the remaining life of the stations would be made by extending the burn-up to 30 GWd/t U at Hunterston B and Torness. There would be additional savings of about Pound 4 million/year in replacement fuel costs if the reactors continued to be refuelled at 30% power at Hunterston B and 40% power at Torness. (author)

  20. Reactivity and isotopic composition of spent PWR [pressurized-water-reactor] fuel as a function of initial enrichment, burnup, and cooling time

    International Nuclear Information System (INIS)

    Cerne, S.P.; Hermann, O.W.; Westfall, R.M.

    1987-10-01

    This study presents the reactivity loss of spent PWR fuel due to burnup in terms of the infinite lattice multiplications factor, k/sub ∞/. Calculations were performed using the SAS2 and CSAS1 control modules of the SCALE system. The k/sub ∞/ values calculated for all combinations of six enrichments, seven burnups, and five cooling times. The results are presented as a primary function of enrichment in both tabular and graphic form. An equation has been developed to estimate the tabulated values of k/sub ∞/'s by specifying enrichment, cooling time, and burnup. Atom densities for fresh fuel, and spent fuel at cooling times of 2, 10, and 20 years are included. 13 refs., 8 figs., 8 tabs

  1. Structural behaviour of fuel assemblies for water cooled reactors. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    2005-07-01

    At the invitation of the Government of France and in response to a proposal of the IAEA Technical Working Group on Water Reactor Fuel Performance and Technology (TWGFPT), the IAEA convened a Technical Meeting on Fuel Assembly Structural Behaviour in Cadarache, France, from 22 to 26 November 2004. The meeting was hosted by the CEA Cadarache Centre, AREVA Framatome-ANP and Electricite de France. The meeting aimed to provide in depth technical exchanges on PWR and WWER operational experience in the field of fuel assembly mechanical behaviour and the potential impact of future high burnup fuel management on fuel reliability. It addressed in-service experience and remedial solutions, loop testing experience, qualification and damage assessment methods (analytic or experimental ones), mechanical behaviour of the fuel assembly including dynamic and fluid structure interaction aspects, modelling and numerical analysis methods, and impact of the in-service evolution of the structural materials. Sixty-seven participants from 17 countries presented 30 papers in the course of four sessions. The topics covered included the impact of hydraulic loadings on fuel assembly (FA)performance, FA bow and control rod (CR) drop kinetics, vibrations and rod-to-grid wear and fretting, and, finally, evaluation and modelling of accident conditions, mainly from seismic causes. FA bow, CR drop kinetics and hydraulics are of great importance under conditions of higher fuel duties including burnup increase, thermal uprates and longer fuel cycles. Vibrations and rod-to-grid wear and fretting have been identified as a key cause of fuel failure at PWRs during the past several years. The meeting demonstrated that full-scale hydraulic tests and modelling provide sufficient information to develop remedies to increase FA skeleton resistance to hydraulic loads, including seismic ones, vibrations and wear. These proceedings are presented as a book with an attached CD-ROM. The first part of the CD

  2. Development and verification of the LIFE-GCFR computer code for predicting gas-cooled fast-reactor fuel-rod performance

    International Nuclear Information System (INIS)

    Hsieh, T.C.; Billone, M.C.; Rest, J.

    1982-03-01

    The fuel-pin modeling code LIFE-GCFR has been developed to predict the thermal, mechanical, and fission-gas behavior of a Gas-Cooled Fast Reactor (GCFR) fuel rod under normal operating conditions. It consists of three major components: thermal, mechanical, and fission-gas analysis. The thermal analysis includes calculations of coolant, cladding, and fuel temperature; fuel densification; pore migration; fuel grain growth; and plenum pressure. Fuel mechanical analysis includes thermal expansion, elasticity, creep, fission-product swelling, hot pressing, cracking, and crack healing of fuel; and thermal expansion, elasticity, creep, and irradiation-induced swelling of cladding. Fission-gas analysis simultaneously treats all major mechanisms thought to influence fission-gas behavior, which include bubble nucleation, resolution, diffusion, migration, and coalescence; temperature and temperature gradients; and fission-gas interaction with structural defects

  3. Evaluation Of Radioactivity Concentration In The Primary Cooling Water System Of The RSG-GAS During Operation With 30% Silicide Fuels

    International Nuclear Information System (INIS)

    Hartoyo, Unggul; Udiyani, P.M.; Setiawanto, Anto

    2001-01-01

    The evaluating radioactivity concentration in the primary cooling water of the RSG-GAS during operation with 30% silicide fuels has been performed. The method of the research is sampling of primary cooling water during operation of the reactor and calculation of its radioactivity concentration. Based on the data obtained from calculation, the identified nuclides in the water are, Mn-56, Sb-124, Sb-122 and Na-24, under the limit of safety value

  4. Fuel-coolant interaction visualization test for in-vessel corium retention external reactor vessel cooling (IVR-ERVC) condition

    Energy Technology Data Exchange (ETDEWEB)

    Na, Young Su; Hong, Seong Ho; Song, Jin Ho; Hong, Seong Wan [Severe Accident and PHWR Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    A visualization test of the fuel-coolant interaction in the Test for Real cOrium Interaction with water (TROI) test facility was carried out. To experimentally simulate the In-Vessel corium Retention (IVR)- External Reactor Vessel Cooling (ERVC) conditions, prototypic corium was released directly into the coolant water without a free fall in a gas phase before making contact with the coolant. Corium (34.39 kg) consisting of uranium oxide and zirconium oxide with a weight ratio of 8:2 was superheated, and 22.54 kg of the 34.39 kg corium was passed through water contained in a transparent interaction vessel. An image of the corium jet behavior in the coolant was taken by a high-speed camera every millisecond. Thermocouple junctions installed in the vertical direction of the coolant were cut sequentially by the falling corium jet. It was clearly observed that the visualization image of the corium jet taken during the fuel-coolant interaction corresponded with the temperature variations in the direction of the falling melt. The corium penetrated through the coolant, and the jet leading edge velocity was 2.0 m/s. Debris smaller than 1 mm was 15% of the total weight of the debris collected after a fuel-coolant interaction test, and the mass median diameter was 2.9 mm.

  5. Effects of fertile blanket on 600 MWth gas-cooled fast reactors: reactor and fuel cycle model

    International Nuclear Information System (INIS)

    Choi, Hang Bok

    2002-07-01

    A physics study has been performed to search for an optimum size of blanket for a 600 MWth gas-cooled fast reactor under fixed fuel and core specifications. The variables considered in this study are the reflector material, reflector thickness and blanket volume. The parametric calculations have shown that a positive breeding gain can be obtained by deploying 8 m 3 natural uranium blanket on the axial and radial boundaries of the core, surrounded by 40 cm Zr 3 Si 2 reflector. However the blanket core has disadvantages compared to the no-blanket core from the viewpoints of fuel fabrication cost and proliferation risk. On the other hand, the no-blanket core has large uncertainties in the possibility of achieving a positive breeding gain. Therefore further studies are recommended for the no-blanket option to improve the breeding gain and achieve a fissile self-sufficient fuel cycle, which is also proliferation-resistant. As an alternative, the blanket option can be considered, that ensures a positive breeding gain

  6. Fuel Summary for Peach Bottom Unit 1 High-Temperature Gas-Cooled Reactor Cores 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Karel I. Kingrey

    2003-04-01

    This fuel summary report contains background and summary information for the Peach Bottom Unit 1, High-Temperature, Gas-Cooled Reactor Cores 1 and 2. This report contains detailed information about the fuel in the two cores, the Peach Bottom Unit 1 operating history, nuclear parameters, physical and chemical characteristics, and shipping and storage canister related data. The data in this document have been compiled from a large number of sources and are not qualified beyond the qualification of the source documents. This report is intended to provide an overview of the existing data pertaining to spent fuel management and point to pertinent reference source documents. For design applications, the original source documentation must be used. While all referenced sources are available as records or controlled documents at the Idaho National Engineering and Environmental Laboratory (INEEL), some of the sources were marked as informal or draft reports. This is noted where applicable. In some instances, source documents are not consistent. Where they are known, this document identifies those instances and provides clarification where possible. However, as stated above, this document has not been independently qualified and such clarifications are only included for information purposes. Some of the information in this summary is available in multiple source documents. An effort has been made to clearly identify at least one record document as the source for the information included in this report.

  7. Effect of cooled EGR on performance and exhaust gas emissions in EFI spark ignition engine fueled by gasoline and wet methanol blends

    Science.gov (United States)

    Rohadi, Heru; Syaiful, Bae, Myung-Whan

    2016-06-01

    Fuel needs, especially the transport sector is still dominated by fossil fuels which are non-renewable. However, oil reserves are very limited. Furthermore, the hazardous components produced by internal combustion engine forces many researchers to consider with alternative fuel which is environmental friendly and renewable sources. Therefore, this study intends to investigate the impact of cooled EGR on the performance and exhaust gas emissions in the gasoline engine fueled by gasoline and wet methanol blends. The percentage of wet methanol blended with gasoline is in the range of 5 to 15% in a volume base. The experiment was performed at the variation of engine speeds from 2500 to 4000 rpm with 500 intervals. The re-circulated exhaust gasses into combustion chamber was 5%. The experiment was performed at the constant engine speed. The results show that the use of cooled EGR with wet methanol of 10% increases the brake torque up to 21.3%. The brake thermal efficiency increases approximately 39.6% using cooled EGR in the case of the engine fueled by 15% wet methanol. Brake specific fuel consumption for the engine using EGR fueled by 10% wet methanol decreases up to 23% at the engine speed of 2500 rpm. The reduction of CO, O2 and HC emissions was found, while CO2 increases.

  8. Economic Viability of Metallic Sodium-Cooled Fast Reactor Fuel in Korea

    Directory of Open Access Journals (Sweden)

    S. K. Kim

    2013-01-01

    Full Text Available This paper evaluates whether SFR metallic nuclear fuel can be economical. To make this determination, the cost of SFCF (SFR fuel cycle facilities was estimated, and the break-even point of the manufacturing cost of SFR metallic nuclear fuel for direct disposal option was then calculated. As a result of the cost estimation, the levelized unit cost (LUC for SFCF was calculated to be 5,311 $/kgHM, and the break-even point was calculated to be $5,267/kgHM. Therefore, the cost difference between LUC and the break-even point is not only small but is also within the relevant range of the uncertainty level of Class 3 in accordance with a generic cost estimate classification matrix of AACE (the Association for the Advancement of Cost Engineering. This means it is very difficult to judge the economical feasibility of SFR metallic nuclear fuel because as of today there are no commercial facilities in Korea or the world. The economic feasibility of SFR metallic nuclear fuel, however, will be enhanced if the mass production of SFCF becomes possible in the future.

  9. Reprocessing of gas-cooled reactor particulate graphite fuel in a multi-strata transmutation system

    International Nuclear Information System (INIS)

    Laidler, J.J.

    2001-01-01

    Spent nuclear fuel discharged for light water reactors (LWRs) contains significant quantities of plutonium and other transuranic elements. Recent practice in Europe and Japan has been to recover the plutonium from spent fuel and recycle it to LWRs in the form of mixed uranium-plutonium oxide (MOX) fuel. Irradiation of the recycle fuel results in the generation of further plutonium and an increase in the isotopic concentration of the higher isotopes of plutonium, those having much lover fission cross sections than 239 Pu. This restricts plutonium recycle to one or two cycles, after which use of the plutonium becomes economically unfavorable. Recycle of the highly-transmuted plutonium in fast spectrum reactors can be an efficient method of fissioning this plutonium as well as other minor transuranics such as neptunium, americium and perhaps even curium. Those minor transuranics that are not conveniently burned in a fast reactor can be sent to an accelerator driven subcritical transmutation device for ultimate destruction. The preceding describes what has become known as a 'dual strata' or 'multi-strata' system. It is driven by the incentives to realize the maximum amount of energy from nuclear fuel and to eliminate the discharge of radio-toxic transuranic elements to the environment. Its implementation will be dependent in the long run upon the economic viability of the system and on the value placed by society on the elimination of radio-toxic materials that can conceivably be used in the manufacture of weapons of mass destruction. (author)

  10. Convective Heat Transfer with and without Film Cooling in High Temperature, Fuel Rich and Lean Environments

    Science.gov (United States)

    2014-09-01

    Figure 3.2: Control panel used in experimental facility the controller throttles a solenoid valve until the desired flow rate is achieved. Finally...test objectives. Namely, quartz windows enabled visual spectrum images or laser diagnostics of reacting film cooling layers. Additionally, a sapphire...rate by a solenoid which throttled a valve restrict the flow appropriately. To measure the flow, the controllers pass the bulk flow through a

  11. The long term storage of advanced gas-cooled reactor (AGR) fuel

    International Nuclear Information System (INIS)

    Standring, P.N.

    1999-01-01

    The approach being taken by BNFL in managing the AGR lifetime spent fuel arisings from British Energy reactors is given. Interim storage for up to 80 years is envisaged for fuel delivered beyond the life of the Thorp reprocessing plant. Adopting a policy of using existing facilities, to comply with the principles of waste minimisation, has defined the development requirements to demonstrate that this approach can be undertaken safely and business issues can be addressed. The major safety issues are the long term integrity of both the fuel being stored and structure it is being stored in. Business related issues reflect long term interactions with the rest of the Sellafield site and storage optimisation. Examples of the development programme in each of these areas is given. (author)

  12. Consequences of pipe ruptures in metal fueled, liquid metal cooled reactors

    International Nuclear Information System (INIS)

    Dunn, F.E.

    1990-01-01

    The capability to simulate pipe ruptures has recently been added to the SASSYS-1 LMR systems analysis code. Using this capability, the consequences of severe pipe ruptures in both loop-type and pool-type reactors using metal fuel were investigated. With metal fuel, if the control rods scram then either type of reactor can easily survive a complete double-ended break of a single pipe; although, as might be expected, the consequences are less severe for a pool-type reactor. A pool-type reactor can even survive a protected simultaneous breaking of all of its inlet pipes without boiling of the coolant or melting of the fuel or cladding. 2 refs., 16 figs., 1 tab

  13. Start-up fuel and power flattening of sodium-cooled candle core

    International Nuclear Information System (INIS)

    Takaki, Naoyuki; Sagawa, Yu; Umino, Akitake; Sekimoto, Hiroshi

    2013-01-01

    The hard neutron spectrum and unique power shape of CANDLE enable its distinctive performances such as achieving high burnup more than 30% and exempting necessity of both enrichment and reprocessing. On the other hand, they also cause several challenging problems. One is how the initial fuel can be prepared to start up the first CANDLE reactor because the equilibrium fuel composition that enables stable CANDLE burning is complex both in axial and radial directions. Another prominent problem is high radial power peaking factor that worsens averaged burnup, namely resource utilization factor in once-through mode and shorten the life time of structure materials. The purposes of this study are to solve these two problems. Several ideas for core configurations and startup fuel using single enrichment uranium and iron as a substitute of fission products are studied. As a result, it is found that low enriched uranium is applicable to ignite the core but all concepts examined here exceeded heat limits. Adjustment in enrichment and height of active and burnt zone is opened for future work. Sodium duct assemblies and thorium fuel assemblies loaded in the center region are studied as measures to reduce radial power peaking factor. Replacing 37 fuels by thorium fuel assemblies in the zeroth to third row provides well-balanced performance with flattened radial power distribution. The CANDLE core loaded with natural uranium in the outer and thorium in the center region achieved 35.6% of averaged burnup and 7.0 years of cladding life time owing to mitigated local fast neutron irradiation at the center. Using thorium with natural or depleted uranium in CANDLE reactor is also beneficial to diversifying fission resource and extending available term of fission energy without expansion of needs for enrichment and reprocessing

  14. The role of a fuel element and its cladding in water cooled reactor dynamics

    International Nuclear Information System (INIS)

    Randles, J.

    1963-10-01

    To clarify the role of fuel element cladding in water reactor dynamics, the heat diffusion and transfer equations are solved in slab geometry for (a) an oscillatory fission power, (b) an oscillatory coolant temperature. From the resulting transfer functions a clear description of the effect of the cladding on the heat flow is obtained. A Mercury autocode programme for evaluating the transfer functions is described. In addition to the slab element, the heat diffusion equations are also solved for a cylindrical system exposed to an oscillatory fission power. The solutions are expressed as transfer functions and are obtainable numerically from another autocode programme. Both of these programmes are used to obtain the power out/ power in transfer function for a typical cylindrical and slab UO 2 fuel pellet clad in zircaloy. The results give a further indication of the effect of the cladding heat capacity over a wide frequency range. It is shown also that the effect of the geometrical difference between a slab and cylindrical fuel element is unimportant provided the surface to volume ratio of the fuel is the same in each case. (author)

  15. The role of a fuel element and its cladding in water cooled reactor dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Randles, J [Technical Assessments and Services Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1963-10-15

    To clarify the role of fuel element cladding in water reactor dynamics, the heat diffusion and transfer equations are solved in slab geometry for (a) an oscillatory fission power, (b) an oscillatory coolant temperature. From the resulting transfer functions a clear description of the effect of the cladding on the heat flow is obtained. A Mercury autocode programme for evaluating the transfer functions is described. In addition to the slab element, the heat diffusion equations are also solved for a cylindrical system exposed to an oscillatory fission power. The solutions are expressed as transfer functions and are obtainable numerically from another autocode programme. Both of these programmes are used to obtain the power out/ power in transfer function for a typical cylindrical and slab UO{sub 2} fuel pellet clad in zircaloy. The results give a further indication of the effect of the cladding heat capacity over a wide frequency range. It is shown also that the effect of the geometrical difference between a slab and cylindrical fuel element is unimportant provided the surface to volume ratio of the fuel is the same in each case. (author)

  16. Development of an Alternative Corrosion Inhibitor for the Storage of Advanced Gas-Cooled Reactor Fuel

    International Nuclear Information System (INIS)

    Standring, P.N.; Hands, B.J.; Morgan, S.; Brooks, A.

    2015-01-01

    Sellafield Lt. currently stores AGR fuel in sodium hyrodxide dosed pool water to pH 11.5 to prevent susceptible AGR fuel from failing due to inter-granular attack. The exception to the above storage practice is Thorp Receipt and Storage (TR&S) where an AGR reprocessing buffer is stored in demineralised water as the expected storage durations were short term (up to 5 years). With the extended shut-down of Thorp, storage durations have increased and this has prompted a re-evaluation of the AGR storage regime in TR&S. The use of sodium hydroxide is not feasible due to a compatibility issue with aluminum components used in LWR storage furniture. The implementation process adopted by Sellafield Ltd in developing an alternative corrosion inhibitor for spent AGR fuel is outlined. The two stranded approach evaluates the impact of candidate corrosion inhibitors on fuel integrity and on plant and processes. The development studies in support of the fuel integrity strand are reported. Candidate inhibitors were first evaluated inactively in terms of their ability to arrest propagating corrosion, radiation stability, compatibility with aluminium and environmental impact. Sodium Nitrate was concluded to be the most promising inhibitor. Sodium nitrate was subsequently tested with active AGR brace material. These studies involved the use of bespoke test equipment and techniques. The studies demonstrated that propagating corrosion could be arrested using 10 ppm nitrate and showed that the resultant nitrate film required relatively high chloride concentrations to break it down over the study duration of 60 days. The development studies to date have provided the confidence that sodium nitrate has the potential to be an effective inhibitor for AGR fuel. The final phase of the fuel integrity strand involves a Lead Container Study using whole AGR pins. A staged approach is being adopted in the study programme where proceeding to a more onerous study is not progressed until positive

  17. Modeling minor actinide multiple recycling in a lead-cooled fast reactor to demonstrate a fuel cycle without long-lived nuclear waste

    Directory of Open Access Journals (Sweden)

    Stanisz Przemysław

    2015-09-01

    Full Text Available The concept of closed nuclear fuel cycle seems to be the most promising options for the efficient usage of the nuclear energy resources. However, it can be implemented only in fast breeder reactors of the IVth generation, which are characterized by the fast neutron spectrum. The lead-cooled fast reactor (LFR was defined and studied on the level of technical design in order to demonstrate its performance and reliability within the European collaboration on ELSY (European Lead-cooled System and LEADER (Lead-cooled European Advanced Demonstration Reactor projects. It has been demonstrated that LFR meets the requirements of the closed nuclear fuel cycle, where plutonium and minor actinides (MA are recycled for reuse, thereby producing no MA waste. In this study, the most promising option was realized when entire Pu + MA material is fully recycled to produce a new batch of fuel without partitioning. This is the concept of a fuel cycle which asymptotically tends to the adiabatic equilibrium, where the concentrations of plutonium and MA at the beginning of the cycle are restored in the subsequent cycle in the combined process of fuel transmutation and cooling, removal of fission products (FPs, and admixture of depleted uranium. In this way, generation of nuclear waste containing radioactive plutonium and MA can be eliminated. The paper shows methodology applied to the LFR equilibrium fuel cycle assessment, which was developed for the Monte Carlo continuous energy burnup (MCB code, equipped with enhanced modules for material processing and fuel handling. The numerical analysis of the reactor core concerns multiple recycling and recovery of long-lived nuclides and their influence on safety parameters. The paper also presents a general concept of the novel IVth generation breeder reactor with equilibrium fuel and its future role in the management of MA.

  18. Review of the SIMMER-II analyses of liquid-metal-cooled fast breeder reactor core-disruptive accident fuel escape

    International Nuclear Information System (INIS)

    DeVault, G.P.; Bell, C.R.

    1985-01-01

    Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. This paper presents a review of analyses with the SIMMER-II computer program of the effectiveness of possible fuel escape paths. Where possible, how SIMMER-II compares with or is validated against experiments that simulated the escape paths also is discussed

  19. A fuselage/tank structure study for actively cooled hypersonic cruise vehicles, summary. [aircraft design of aircraft fuel systems

    Science.gov (United States)

    Pirrello, C. J.; Baker, A. H.; Stone, J. E.

    1976-01-01

    A detailed analytical study was made to investigate the effects of fuselage cross section (circular and elliptical) and the structural arrangement (integral and nonintegral tanks) on aircraft performance. The vehicle was a 200 passenger, liquid hydrogen fueled Mach 6 transport designed to meet a range goal of 9.26 Mn (5000 NM). A variety of trade studies were conducted in the area of configuration arrangement, structural design, and active cooling design in order to maximize the performance of each of three point design aircraft: (1) circular wing-body with nonintegral tanks, (2) circular wing-body with integral tanks and (3) elliptical blended wing-body with integral tanks. Aircraft range and weight were used as the basis for comparison. The resulting design and performance characteristics show that the blended body integral tank aircraft weights the least and has the greatest range capability, however, producibility and maintainability factors favor nonintegral tank concepts.

  20. Method of detecting stacks with leaky fuel elements in liquid-metal-cooled reactor and apparatus for effecting same

    International Nuclear Information System (INIS)

    Aristarkhov, N.N.; Efimov, I.A.; Zaistev, B.I.; Peters, I.G.; Tymosh, B.S.

    1976-01-01

    Described is a method of detecting stacks with leaky fuel elements in a liquid-metal-cooled reactor, consisting in that prior to withdrawing a coolant sample, gas is accumulated in the coolant of the stack being controlled, the reactor being shut down, separated from the sample by means of an inert carrier gas, and the radioactivity of the separated gas is measured. An apparatus for carrying out said method comprises a sampler in the form of a tube parallel to the reactor axis in the hole of a rotating plug and adapted to move along the reactor axis. Made in the top portion of the tube are holes for the introduction of the inert carrier gas and the removal thereof together with the gases evolved from the coolant, while the bottom portion of the tube is provided with a sealing member

  1. Calculation of mass flow and steam quality distribution on fuel elements of light-water cooled boiling water nuclear reactors

    International Nuclear Information System (INIS)

    Hermanns, H.J.

    1977-04-01

    By the example of light-water cooled nuclear reactors, the state of the calculation methods at disposal for calculating mass flow and steam quality distribution (sub-channel analysis) is indicated. Particular regard was paid to the transport phenomena occurring in reactor fuel elements in the range of two phase flow. Experimentally determined values were compared with recalculations of these experiments with the sub-channel code COBRA; from the results of these comparing calculations, conclusions could be drawn on the suitability of this code for defined applications. Limits of reliability could be determined to some extent. Based on the experience gained and the study of individual physical model concepts, recognized as being important, a sub-channel model was drawn up and the corresponding numerical computer code (SIEWAS) worked out. Experiments made at GE could be reproduced with the code SIEWAS with sufficient accuracy. (orig.) [de

  2. Status of sodium cooled fast reactors with closed fuel cycle in India

    International Nuclear Information System (INIS)

    Raj, B.

    2007-01-01

    Fast reactors form the second stage of India's 3-stage nuclear power programme. The seed for India's fast reactor programme was sown through the construction of the Fast Breeder Test Reactor (FBTR) at IGCAR, Kalpakkam, that was commissioned in 1985. FBTR has operated with an unique, indigenously developed plutonium rich mixed carbide fuel, which has reached a burn up as high as 155 GWd/t without any fuel failure in the core. The sodium systems in the reactor have performed excellently. The availability of the reactor has been as high as 92% in the recent campaigns. The fuel discharged from FBTR up to 100 GWd/t has been reprocessed successfully. The experience gained in the construction, commissioning and operation of FBTR has provided the necessary confidence to launch a Prototype FBR of 500 MWe capacity (PFBR). This reactor will be fuelled by uranium, plutonium mixed oxide. The reactor construction started in 2003 and the reactor is scheduled to be commissioned by 2010. The design of the reactor has incorporated the worldwide operating experience from the FBRs and has addressed various safety issues reported in literature, besides introducing a number of innovative features which have reduced the unit energy cost and contributed to its enhanced safety. Simultaneous with the construction of the reactor, the fuel cycle of the reactor has been addressed in a comprehensive manner and construction of a fuel cycle facility has been initiated. Subsequent to the PFBR, 4 more reactors with identical design are proposed to be constructed. Various elements of reactor design are being carefully analysed with the aim of introducing innovative features towards further reduction in unit energy cost and enhancing safety in these reactors

  3. Cooling Performance Characteristics of the Stack Thermal Management System for Fuel Cell Electric Vehicles under Actual Driving Conditions

    Directory of Open Access Journals (Sweden)

    Ho-Seong Lee

    2016-04-01

    Full Text Available The cooling performance of the stack radiator of a fuel cell electric vehicle was evaluated under various actual road driving conditions, such as highway and uphill travel. The thermal stability was then optimized, thereby ensuring stable operation of the stack thermal management system. The coolant inlet temperature of the radiator in the highway mode was lower than that associated with the uphill mode because the corresponding frontal air velocity was higher than obtained in the uphill mode. In both the highway and uphill modes, the coolant temperatures of the radiator, operated under actual road driving conditions, were lower than the allowable limit (80 °C; this is the maximum temperature at which stable operation of the stack thermal management system of the fuel cell electric vehicle could be maintained. Furthermore, under actual road driving conditions in uphill mode, the initial temperature difference (ITD between the coolant temperature and air temperature of the system was higher than that associated with the highway mode; this higher ITD occurred even though the thermal load of the system in uphill mode was greater than that corresponding to the highway mode. Since the coolant inlet temperature is expected to exceed the allowable limit (80 °C in uphill mode under higher ambient temperature with air conditioning system operation, the FEM design layout should be modified to improve the heat capacity. In addition, the overall volume of the stack cooling radiator is 52.2% higher than that of the present model and the coolant inlet temperature of the improved radiator is 22.7% lower than that of the present model.

  4. Aseismic study of high temperature gas-cooled reactor core with block-type fuel, 3

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1985-01-01

    A two-dimensional horizontal seismic experiment with single axis and simultaneous two-axes excitations was performed to obtain the core seismic design data on the block-type high temperature gas-cooled reactor. Effects of excitation directions and core side support stiffness on characteristics of core displacements and reaction forces of support were revealed. The values of the side reaction forces are the largest in the excitation of flat-to-flat of hexagonal block. Preload from the core periphery to the core center are effective to decrease core displacements and side reaction forces. (author)

  5. Alternative Liquid Fuel Effects on Cooled Silicon Nitride Marine Gas Turbine Airfoils

    Energy Technology Data Exchange (ETDEWEB)

    Holowczak, J.

    2002-03-01

    With prior support from the Office of Naval Research, DARPA, and U.S. Department of Energy, United Technologies is developing and engine environment testing what we believe to be the first internally cooled silicon nitride ceramic turbine vane in the United States. The vanes are being developed for the FT8, an aeroderivative stationary/marine gas turbine. The current effort resulted in further manufacturing and development and prototyping by two U.S. based gas turbine grade silicon nitride component manufacturers, preliminary development of both alumina, and YTRIA based environmental barrier coatings (EBC's) and testing or ceramic vanes with an EBC coating.

  6. Potentials of cooled EGR and water injection for knock resistance and fuel consumption improvements of gasoline engines

    International Nuclear Information System (INIS)

    Bozza, Fabio; De Bellis, Vincenzo; Teodosio, Luigi

    2016-01-01

    Highlights: • 1D simulation of a turbocharged VVA engine under knock limited operation. • Description of turbulence, combustion and knock by phenomenological models. • Comparison of EGR and ported water injection at high load for knock mitigation and fuel economy. • Virtual calibration of engine control parameters by a 1D model. - Abstract: It is well known that the downsizing philosophy allows the improvement of the brake specific fuel consumption (BSFC) at part load operation for spark ignition (SI) engines. On the other hand, the BSFC is penalized at high load because of the knock occurrence and of further limitations on the turbine inlet temperature (TIT). Knock control forces the adoption of a late combustion phasing, causing a deterioration of the thermodynamic efficiency, while the TIT control requires the enrichment of the air-to-fuel ratio (A/F), with additional BSFC drawbacks. In this work, two promising techniques are investigated by a 1D approach with the aim of improving the fuel economy of a turbocharged SI engine at full load knock-limited operation. The first technique is the recirculation of low-pressure cooled exhaust gas (EGR), while the second is the injection of liquid water at the intake ports. Proper “in-house developed” sub-models are used to describe the turbulence, combustion and knock phenomena. The effects of the above techniques are studied in six operating points at full load and different speeds for various A/F levels and inert content, by varying the EGR rate and water-to-fuel ratio. The presented results highlight that both the solutions involve significant BSFC improvements, especially in the operating conditions at medium engine speeds. In fact, the introduction of inert gas in the cylinder contributes to reduce the knock tendency, resulting in the possibility to advance the combustion phasing and reduce, or even avoid, the mixture over-fuelling. The heat subtracted by the water evaporation enhances the above effects

  7. Failed fuel monitoring and surveillance techniques for liquid metal cooled fast reactors

    International Nuclear Information System (INIS)

    Lambert, J.D.B.; Mikaili, R.; Gross, K.C.; Strain, R.V.; Aoyama, T.; Ukai, S.; Nomura, S.; Nakae, N.

    1995-01-01

    The Experimental Breeder Reactor II (EBR-II) has been used as a facility for irradiation of LMR fuels and components for thirty years. During this time many tests of experimental fuel were continued to cladding breach in order to study modes of element failure; the methods used to identify such failures are described in a parallel paper. This paper summarizes experience of monitoring the delayed-neutron (DN) and fission-gas (FG) release behavior of a smaller number of elements that continued operation in the run-beyond-cladding-breach (RBCB) mode. The scope of RBCB testing, the methods developed to characterize failures on-line, and examples of DN/FG behavior are described

  8. LWR fuel cladding deformation in a LOCA and its interaction with the emergency core cooling

    International Nuclear Information System (INIS)

    Erbacher, F.J.

    1982-01-01

    The paper summarizes research results of out-of-pile burst tests, in-pile bursts tests, out-of-pile flooding tests and modeling work on fuel behavior in a LOCA performed at KfK: The dominant phenomena of the cladding deformation and failure have been clarified by experiments and can be modeled by computer codes. The burst and flooding tests performed up to now suggest that the coolability of the core under LOCA conditions can be maintained. (orig.) [de

  9. Zircaloy PWR fuel cladding deformation tests under mainly convective cooling conditions

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.

    1980-01-01

    In a loss-of-coolant accident the temperature of the cladding of the fuel rods may rise to levels (650-810 0 C) where the ductility of Zircaloy is high (approximately 80%). The net outward pressure which will obtain if the coolant pressure falls to a small fraction of its normal working value produces stresses in the cladding which can result in large strain through secondary creep. An earlier study of the deformation of specimens of PWR Zircaloy cladding tubing 450 mm long under internal pressure had shown that strains of over 50% could be produced over considerable lengths (greater than twenty tube diameters). Extended deformation of this sort might be unacceptable if it occurred in a fuel element. The previous tests had been carried out under conditions of uniform radiative heat loss, and the work reported here extends the study to conditions of mainly convective heat loss believed to be more representative of a fuel element following a loss of coolant. Zircaloy-4 cladding specimens 450 mm long were filled with alumina pellets and tested at temperatures between 630 and 845 0 C in flowing steam at atmospheric pressure. Internal test pressures were in the range 2.9-11.0 MPa (400-1600 1b/in 2 ). Maximum strains were observed of the same magnitude as those seen in the previous tests, but the shape of the deformation differed; in these tests the deformation progressively increased in the direction of the steam flow. These results are compared with those from multi-rod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behaviour of fuel elements in a loss-of-coolant accident are outlined. (author)

  10. The use of low enriched uranium fuel cycle in high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    The present paper begins with a brief review of the status of research and development of experimental VHTR in Japan. On the basis of the experience gained from these work, assessment is made of commercial HTRs. Material balance with fuel burnup is calculated for the two core models; one is HTGR for steam cycle and the other VHTR for process heat application. The results of assessment of commercial HTRs are compared with those for LWR

  11. Operation of Atucha I nuclear power plant with 25 cooling channels without fuel elements

    International Nuclear Information System (INIS)

    Perez, R.A.; Sidelnik, J.I.; Salom, G.F.

    1987-01-01

    In view of the need of removing the irradiation probes from the reactor of Atucha I nuclear power plant, a study about the consequences of operating with 25 channels without their respective fuel elements was performed. This condition was simulated by means of the code PUMA symmetry I and the consequences were analyzed. From the study resulted a program of stepped power reduction of the nuclear plant that would take place during the process of channel emptying. (Author)

  12. Operator decision aid for breached fuel operation in liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Gross, K.C.; Hawkins, R.E.; Nickless, W.K.

    1991-01-01

    The purpose of this paper is to report the development of an expert system that provides continuous assessment of the safety significance and technical specification conformance of Delayed Neutron (DN) signals during breached fuel operation. The completed expert system has been parallelized on an innovative distributed-memory network-computing system that enables the computationally intensive kernel of the expert system to run in parallel on a group of low-cost Unix workstations. 1 ref

  13. Characteristic Evaluation on the Cooling Performance of an Electrical Air Conditioning System Using R744 for a Fuel Cell Electric Vehicle

    Directory of Open Access Journals (Sweden)

    Moo-Yeon Lee

    2012-05-01

    Full Text Available The objective of this study was to investigate the cooling performance characteristics of an electrical air conditioning system using R744 as an alternative of R-134a for a fuel cell electric vehicle. In order to analyze the cooling performance characteristics of the air conditioning system using R744 for a fuel cell electric vehicle, an electrical air conditioning system using R744 was developed and tested under various operating conditions according to both inlet air conditions of the gas cooler and evaporator and compressor speed. The cooling capacity and coefficient of performance (COP forcooling of the tested air conditioning system were up to 6.4 kW and 2.5, respectively. In addition, the electrical air conditioning system with R744 using an inverter driven compressor showed better performance than the conventional air conditioning system with R-134a under the same operating conditions. The observed cooling performance of the developed electrical air conditioning system was found to be sufficient for cooling loads under various real driving conditions for a fuel cell electric vehicle.

  14. Theoretical investigations of the meltoff and resolidification process of fuel claddings during accidents in liquid metal cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Angerer, G.

    1978-08-01

    During loss-of-coolant-flow accidents in liquid metal cooled fast breeder reactors with failure to scram the fuel claddings will melt after boiling and evaporation of the coolant. The CMOT model presented here describes the subsequent process of relocation and resolidification of the molten claddings. The basic thermohydrodynamics equations of the two-phase flow of cladding material and sodium vapor are solved numerically by differential approximations in a Eulerian reference net. The results calculated by the model improved the insight into the dynamics of the cladding relocation process. Here are the main results: - Shortly after the onset of cladding relocation large waves of molten cladding material are generated. The motion of these waves contributes considerably to the material transport. - The dynamics of cladding relocation exhibits strong local incoherences. - The formation of cladding blockages observed at the ends of the fuel region is confirmed by the calculations. - In case of incoherent cladding meltoff less cladding material is transported upwards. - Cladding relocation strongly depends on the axial pressure drop and the underlying friction factor correlations. Recalculation of the R5 loss-of-coolant-flow experiment performed in the U.S. TREAT test reactor is in good agreement with the experimental data. (orig./HP) 891 HP [de

  15. Study on MAs transmutation of accelerator-driven system sodium-cooled fast reactor loaded with metallic fuel

    International Nuclear Information System (INIS)

    Han Song; Yang Yongwei

    2007-01-01

    Through the analysis of the effect of heavy metal actinides on the effective multiplication constant (k eff ) of the core in accelerator-driven system (ADS) sodium-cooled fast reactor loaded with metallic fuel, we gave the method for determining fuel components. the characteristics of minor actinides (MAs) transmutation was analyzed in detail. 3D burn-up code COUPLE, which couples MCNP4c3 and ORIGEN2, was applied to the neutron simulation and burn up calculation. The results of optimized scheme shows that adjusting the proportion of 239 Pu and maintaining the value during the burn-up cycle is an efficient method of designing k eff and keeping stable during the burn-up cycle. Spallation neutrons lead to the neutron spectrum harder at inner core than that at outer core. It is in favor of improving MA's fission cross sections and the capture-to-fission ratio. The total MAs transmutation support ratio 8.3 achieves excellent transmutation effect. For higher flux at inner core leads to obvious differences on transmutation efficiency,only disposing MAs at inner core is in favor of decreasing the loading mass and improving MAs transmutation effect. (authors)

  16. Comparison of lead and sodium-cooled reactors - Safety, fuel cycle performance and some economical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Carlsson, Johan; Tucek, Kamil; Wider, Hartmut [Joint Research Centre, EC-JRC, Westerduinweg 3, P.O. Box 2, NL-0 1755 ZG Petten (Netherlands)

    2006-07-01

    This paper compares the Lead-cooled Fast Reactor (LFR) and the Sodium-cooled Fast Reactor (SFR) regarding different aspects of the coolant, safety and economics. A brief review of design and safety experience of an SFR (BN-600) and some safety philosophy of the most developed LFR (BREST) are presented as well. The pros and cons of the lead and the sodium coolants are discussed. This paper presents results concerning the coolant temperature evolution during three accident scenarios, i.e. Loss-Of- Flow (LOF), Loss-Of-Heat-Sink (LOHS), and Total-Loss-Of-Power (TLOP). It also studies possible moderators, like BeO and hydrides, for the core designs to have negative reactivity feedbacks and favorable reactivity swings. LFR seems to be able to accommodate more minor actinides than SFR at comparable coolant and Doppler feedbacks. We show that LFR can be designed both to breed and burn transuranics from LWRs. The hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. It is shown that the LFR can handle the LOF transient better than the SFR. This is due to the much lower pressure drop in the LFR core. The coolant outlet temperatures stabilize at 2050 K and 940 K for SFR and LFR, respectively when no feedbacks are considered. Investigations also concern the SFR's performance when the pitch-to-diameter was increased from 1.2 to 1.4. For the LOHS and TLOP accidents their temperature evolutions are milder for the LFR since lead has a 50% larger volumetric heat capacity. For the TLOP the core outlet temperature of the LFR peaks at 1080 K after 2 days. Regarding economics it appears easier to avoid an intermediate cycle in an LFR than an SFR. (authors)

  17. Comparison of lead and sodium-cooled reactors - Safety, fuel cycle performance and some economical aspects

    International Nuclear Information System (INIS)

    Carlsson, Johan; Tucek, Kamil; Wider, Hartmut

    2006-01-01

    This paper compares the Lead-cooled Fast Reactor (LFR) and the Sodium-cooled Fast Reactor (SFR) regarding different aspects of the coolant, safety and economics. A brief review of design and safety experience of an SFR (BN-600) and some safety philosophy of the most developed LFR (BREST) are presented as well. The pros and cons of the lead and the sodium coolants are discussed. This paper presents results concerning the coolant temperature evolution during three accident scenarios, i.e. Loss-Of- Flow (LOF), Loss-Of-Heat-Sink (LOHS), and Total-Loss-Of-Power (TLOP). It also studies possible moderators, like BeO and hydrides, for the core designs to have negative reactivity feedbacks and favorable reactivity swings. LFR seems to be able to accommodate more minor actinides than SFR at comparable coolant and Doppler feedbacks. We show that LFR can be designed both to breed and burn transuranics from LWRs. The hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. It is shown that the LFR can handle the LOF transient better than the SFR. This is due to the much lower pressure drop in the LFR core. The coolant outlet temperatures stabilize at 2050 K and 940 K for SFR and LFR, respectively when no feedbacks are considered. Investigations also concern the SFR's performance when the pitch-to-diameter was increased from 1.2 to 1.4. For the LOHS and TLOP accidents their temperature evolutions are milder for the LFR since lead has a 50% larger volumetric heat capacity. For the TLOP the core outlet temperature of the LFR peaks at 1080 K after 2 days. Regarding economics it appears easier to avoid an intermediate cycle in an LFR than an SFR. (authors)

  18. Investigation of velocity distribution in an inner subchannel of wire wrapped fuel pin bundle of sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Nishimura, Masahiro; Kamide, Hideki; Ohshima, Hiroyuki; Kobayashi, Jun; Sato, Hiroyuki

    2011-01-01

    A sodium cooled fast reactor is designed to attain a high burn-up of core fuel in commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, deformation of fuel pin due to the swelling and thermal bowing may decrease local flow velocity via change of flow area in the subassembly and influence the heat removal capability. Therefore, it is important to obtain the detail of flow velocity distribution in a wire wrapped pin bundle. In this study, water experiments were carried out to investigate the detailed velocity distribution in a subchannel of nominal pin geometry as the first step. These basic data are not only useful for understanding of pin bundle thermal hydraulics but also a code validation. A wire-wrapped 3-pin bundle water model was applied to investigate the detailed velocity distribution in the subchannel which is surrounded by 3 pins with wrapping wire. The test section consists of an irregular hexagonal acrylic duct tube and three pins made of fluorinated resin pins which has nearly the same refractive index with that of water and a high light transmission rate. This enables to visualize the central subchannel through the pins. The velocity distribution in the central subchannel with the wrapping wire was measured by PIV (Particle Image Velocimetry) through a side wall of the duct tube. Typical flow velocity conditions in the pin bundle were 0.36m/s (Re=2,700) and 1.6m/s (Re=13,500). Influence of the wrapping wire on the velocity distributions in vertical and horizontal directions was confirmed. A clockwise swirl flow around the wire was found in subchannel. Significant differences were not recognized between the two cases of Re=2,700 and 13,500 concerning flow patterns. (author)

  19. Studies on natural circulation cooling enhancement in a spent fuel in fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Isamu; Akamatsu, Mikio; Toda, Shinichi; Sato, Manabu [Kawasaki Heavy Industries Ltd., Kobe (Japan); Mayumi, Masami

    2001-01-01

    Fast breeder reactor (FBR) has some advantages such as effective application of plutonium, excellent capacity to fire minor-actinides (longer half-life nuclides such as Np, Am, Cm, and so on) contained in radioactive wastes in the reactor to convert their shorter half-life nuclides. However, fuels containing the minor-actinides have a characteristic with higher exotherm and radioactive intensity than those of conventional ones, it is essential at their actual stages to prepare some rational fuel handling systems on their transportation, storage and so forth. In addition, there are few examples on natural circulation heat transfer test of a liquid metal using long sized container. Then, in order to establish an evaluating method on decay-heat removing property of a spent fuel assembly in sodium canister and pot, some natural circulation tests on a long sized container including a quasi pin-bundle structure for a working fluid of lead-bismuth (Pb-Bi) mixture with easier handling than that of sodium was carried out. A specimen could be mounted at optional angles from horizontal to vertical positions so as to evaluate effects of inclined angles. In addition, in order to estimate temperature and flow rate distribution in a long sized container and understand thermal flowing phenomenon in specimen system, numerical analysis using multi-dimensional analysis code was carried out. As a result, it was found that in vertical arrangement system, natural circulation phenomenon is limited at upper portion of the exothermal portion, and its maximum temperature was tested at central portion of top pin-bundle of the exothermal portion. And, it was also found that at horizontal arrangement maximum temperature was 40 centigrade less than that of vertical arrangement, and so forth. (G.K.)

  20. An electrical simulator of a nuclear fuel rod cooled by nucleate boiling

    International Nuclear Information System (INIS)

    Costa, Antonio Carlos Lopes da; Machado, Luiz; Koury, Ricardo Nicolau Nassar; Passos, Julio Cesar

    2009-01-01

    This study investigates an electrical heated test section designed to simulate a nuclear fuel rod. This simulator comprises a stainless steel vertical tube, with length and outside diameter of 600 mm and 10 mm, respectively, inside which there is a high power electrical resistor. The heat generated is removed by means of enhanced confined subcooled nucleate boiling of water in an annular space containing 153 small metal inclined discs. The tests were performed under electrical power and pressure up to 48 kW and 40 bar, respectively. The results show that the experimental boiling heat transfer coefficients are in good agreement with those calculated using the Jens-Lottes correlation. (author)

  1. An electrical simulator of a nuclear fuel rod cooled by nucleate boiling

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Antonio Carlos Lopes da [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)], e-mail: aclc@cdtn.br; Machado, Luiz; Koury, Ricardo Nicolau Nassar [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Mecanica], e-mail: luizm@demec.ufmg.br; Bonjour, Jocelyn [CETHIL, UMR5008, CNRS, INSA-Lyon (France)], e-mail: jocelyn.bonjour@insa-lyon.fr; Passos, Julio Cesar [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil). Dept. de Engenharia Mecanica. LEPTEN/Boiling], e-mail: jpassos@emc.ufsc.br

    2009-07-01

    This study investigates an electrical heated test section designed to simulate a nuclear fuel rod. This simulator comprises a stainless steel vertical tube, with length and outside diameter of 600 mm and 10 mm, respectively, inside which there is a high power electrical resistor. The heat generated is removed by means of enhanced confined subcooled nucleate boiling of water in an annular space containing 153 small metal inclined discs. The tests were performed under electrical power and pressure up to 48 kW and 40 bar, respectively. The results show that the experimental boiling heat transfer coefficients are in good agreement with those calculated using the Jens-Lottes correlation. (author)

  2. A Feasibility Study to Determine Cooling Time and Burnup of ATR Fuel Using a Nondestructive Technique and Three Types of Gamma-ray Detectors

    International Nuclear Information System (INIS)

    Navarro, Jorge; Aryaeinejad, Rahmat; Nigg, David W.

    2011-01-01

    Effective and efficient Advanced Test Reactor (ATR) fuel management require state of the art core modeling tools. These new tools will need isotopic and burnup validation data before they are put into production. To create isotopic, burn up validation libraries and to determine the setup for permanent fuel scanner system a feasibility study was perform. The study consisted in measuring short and long cooling time fuel elements at the ATR canal. Three gamma spectroscopy detectors (HPGe, LaBr3, and HPXe) and two system configurations (above and under water) were used in the feasibility study. The first stage of the study was to investigate which detector and system configuration would be better suited for different scenarios. The second stage of the feasibility study was to create burnup and cooling time calibrations using experimental isotopic data collected and ORIGEN 2.2 burnup data. The results of the study establish that a better spectra resolution is achieve with an above the water configuration and that three detectors can be used in the permanent fuel scanner system for different situations. In addition it was conclude that a number of isotopic ratios and absolute measurements could be used to predict ATR fuel burnup and cooling times.

  3. BN800: The advanced sodium cooled fast reactor plant based on close fuel cycle

    International Nuclear Information System (INIS)

    Wu Xingman

    2011-01-01

    As one of the advanced countries with actually fastest reactor technology, Russia has always taken a leading role in the forefront of the development of fast reactor technology. After successful operation of BN600 fast reactor nuclear power station with a capacity of six hundred thousand kilowatts of electric power for nearly 30 years, and after a few decades of several design optimization improved and completed on its basis, it is finally decided to build Unit 4 of Beloyarsk nuclear power station (BN800 fast reactor power station). The BN800 fast reactor nuclear power station is considered to be the project of the world's most advanced fast reactor nuclear power being put into implementation. The fast reactor technology in China has been developed for decades. With the Chinese pilot fast reactor to be put into operation soon, the Chinese model fast reactor power station has been put on the agenda. Meanwhile, the closed fuel cycle development strategy with fast reactor as key aspect has given rise to the concern of experts and decision-making level in relevant areas. Based on the experiences accumulated in many years in dealing the Sino-Russian cooperation in fast reactor technology, with reference to the latest Russian published and authoritative literatures regarding BN800 fast reactor nuclear power station, the author compiled this article into a comprehensive introduction for reference by leaders and experts dealing in the related fields of nuclear fuel cycle strategy and fast reactor technology development researches, etc. (authors)

  4. Experimental study of the deformation of Zircaloy PWR fuel rod cladding under mainly convective cooling

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.

    1982-01-01

    Zircaloy-4 cladding specimens 450 mm long were filled with alumina pellets and tested at temperatures between 630 and 915 degree C in flowing steam at atmospheric pressure. Internal test pressures were in the range 0.69 to 11.0 MPa. The length of cladding strained 33 percent or more was greatest (about 20 times the original diameter) when the initial pressure was 1.38/plus or minus/0.17MPa. This results from oxidation strengthening of the surface layers acting as an additional mechanism for stabilizing the deformation or partial superplastic deformation, or both. For adjacent rods in a fuel assembly not to touch at any temperature, the pressure would have to be less than about 1 MPa. These results are compared with those form multirod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behavior of fuel elements in a loss-of-coolant accident are outlined. 37 refs

  5. Experimental study of the deformation of Zircaloy PWR fuel rod cladding under mainly convective cooling

    Energy Technology Data Exchange (ETDEWEB)

    Hindle, E.D.; Mann, C.A.

    1982-01-01

    Zircaloy-4 cladding specimens 450 mm long were filled with alumina pellets and tested at temperatures between 630 and 915 degree C in flowing steam at atmospheric pressure. Internal test pressures were in the range 0.69 to 11.0 MPa. The length of cladding strained 33 percent or more was greatest (about 20 times the original diameter) when the initial pressure was 1.38/plus or minus/0.17MPa. This results from oxidation strengthening of the surface layers acting as an additional mechanism for stabilizing the deformation or partial superplastic deformation, or both. For adjacent rods in a fuel assembly not to touch at any temperature, the pressure would have to be less than about 1 MPa. These results are compared with those form multirod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behavior of fuel elements in a loss-of-coolant accident are outlined. 37 refs.

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

    International Nuclear Information System (INIS)

    Saxena, Aakanksha; Cadiou, Thierry; Bieder, Ulrich; Viazzo, Stephane

    2013-06-01

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

  7. Mississippi State University Cooling, Heating, and Power (Micro-CHP) and Bio-Fuel Center

    Energy Technology Data Exchange (ETDEWEB)

    Mago, Pedro [Mississippi State Univ., Mississippi State, MS (United States); Newell, LeLe [Mississippi State Univ., Mississippi State, MS (United States)

    2014-01-31

    Between 2008 and 2014, the U.S. Department of Energy funded the MSU Micro-CHP and Bio-Fuel Center located at Mississippi State University. The overall objective of this project was to enable micro-CHP (micro-combined heat and power) utilization, to facilitate and promote the use of CHP systems and to educate architects, engineers, and agricultural producers and scientists on the benefits of CHP systems. Therefore, the work of the Center focused on the three areas: CHP system modeling and optimization, outreach, and research. In general, the results obtained from this project demonstrated that CHP systems are attractive because they can provide energy, environmental, and economic benefits. Some of these benefits include the potential to reduce operational cost, carbon dioxide emissions, primary energy consumption, and power reliability during electric grid disruptions. The knowledge disseminated in numerous journal and conference papers from the outcomes of this project is beneficial to engineers, architects, agricultural producers, scientists and the public in general who are interested in CHP technology and applications. In addition, more than 48 graduate students and 23 undergraduate students, benefited from the training and research performed in the MSU Micro-CHP and Bio-Fuel Center.

  8. Detailed Reaction Kinetics for CFD Modeling of Nuclear Fuel Pellet Coating for High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Battaglia, Francine

    2008-01-01

    The research project was related to the Advanced Fuel Cycle Initiative and was in direct alignment with advancing knowledge in the area of Nuclear Fuel Development related to the use of TRISO fuels for high-temperature reactors. The importance of properly coating nuclear fuel pellets received a renewed interest for the safe production of nuclear power to help meet the energy requirements of the United States. High-temperature gas-cooled nuclear reactors use fuel in the form of coated uranium particles, and it is the coating process that was of importance to this project. The coating process requires four coating layers to retain radioactive fission products from escaping into the environment. The first layer consists of porous carbon and serves as a buffer layer to attenuate the fission and accommodate the fuel kernel swelling. The second (inner) layer is of pyrocarbon and provides protection from fission products and supports the third layer, which is silicon carbide. The final (outer) layer is also pyrocarbon and provides a bonding surface and protective barrier for the entire pellet. The coating procedures for the silicon carbide and the outer pyrocarbon layers require knowledge of the detailed kinetics of the reaction processes in the gas phase and at the surfaces where the particles interact with the reactor walls. The intent of this project was to acquire detailed information on the reaction kinetics for the chemical vapor deposition (CVD) of carbon and silicon carbine on uranium fuel pellets, including the location of transition state structures, evaluation of the associated activation energies, and the use of these activation energies in the prediction of reaction rate constants. After the detailed reaction kinetics were determined, the reactions were implemented and tested in a computational fluid dynamics model, MFIX. The intention was to find a reduced mechanism set to reduce the computational time for a simulation, while still providing accurate results

  9. Heat transfer performance of heat pipe for passive cooling of spent fuel pool

    International Nuclear Information System (INIS)

    Wang Minglu; Xiong Zhengqin; Gu Hanyang; Ye Cheng; Cheng Xu

    2014-01-01

    A large-scale loop heat pipe has no electricity driven component and high efficiency of heat transfer. It can be used for the passive cooling of the SFP after SBO to improve the safety performance of nuclear power plants. In this paper, such a large-scale loop heat pipe is studied experimentally. The heat transfer rate, evaporator average heat transfer coefficient operating temperature, operating pressure and ammonia flow rate have been obtained with the water flow ranging from 0.007 m/s to 0.02 m/s outside the evaporator section, heating water temperature in the range of 50 to 90℃, air velocity outside the condensation section ranging from 0.5 to 2.5 m/s. It is found that the heat transfer rate reaches as high as 20.1 kW. Parametric analysis indicates that, the heat transfer rate and ammonia flow rate are influenced significantly by hot water inlet temperature and velocity, while beyond 1.5 m/s, the effect of air velocity outside the condensation section is minor. (authors)

  10. Benchmark for Neutronic Analysis of Sodium-cooled Fast Reactor Cores with Various Fuel Types and Core Sizes

    International Nuclear Information System (INIS)

    Stauff, N.E.; Kim, T.K.; Taiwo, T.A.; Buiron, L.; Rimpault, G.; Brun, E.; Lee, Y.K.; Pataki, I.; Kereszturi, A.; Tota, A.; Parisi, C.; Fridman, E.; Guilliard, N.; Kugo, T.; Sugino, K.; Uematsu, M.M.; Ponomarev, A.; Messaoudi, N.; Lin Tan, R.; Kozlowski, T.; Bernnat, W.; Blanchet, D.; Brun, E.; Buiron, L.; Fridman, E.; Guilliard, N.; Kereszturi, A.; Kim, T.K.; Kozlowski, T.; Kugo, T.; Lee, Y.K.; Lin Tan, R.; Messaoudi, N.; Parisi, C.; Pataki, I.; Ponomarev, A.; Rimpault, G.; Stauff, N.E.; Sugino, K.; Taiwo, T.A.; Tota, A.; Uematsu, M.M.; Monti, S.; Yamaji, A.; Nakahara, Y.; Gulliford, J.

    2016-01-01

    One of the foremost Generation IV International Forum (GIF) objectives is to design nuclear reactor cores that can passively avoid damage of the reactor when control rods fail to scram in response to postulated accident initiators (e.g. inadvertent reactivity insertion or loss of coolant flow). The analysis of such unprotected transients depends primarily on the physical properties of the fuel and the reactivity feedback coefficients of the core. Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS), the Sodium Fast Reactor core Feed-back and Transient response (SFR-FT) Task Force was proposed to evaluate core performance characteristics of several Generation IV Sodium-cooled Fast Reactor (SFR) concepts. A set of four numerical benchmark cases was initially developed with different core sizes and fuel types in order to perform neutronic characterisation, evaluation of the feedback coefficients and transient calculations. Two 'large' SFR core designs were proposed by CEA: those generate 3 600 MW(th) and employ oxide and carbide fuel technologies. Two 'medium' SFR core designs proposed by ANL complete the set. These medium SFR cores generate 1 000 MW(th) and employ oxide and metallic fuel technologies. The present report summarises the results obtained by the WPRS for the neutronic characterisation benchmark exercise proposed. The benchmark definition is detailed in Chapter 2. Eleven institutions contributed to this benchmark: Argonne National Laboratory (ANL), Commissariat a l'energie atomique et aux energies alternatives (CEA of Cadarache), Commissariat a l'energie atomique et aux energies alternatives (CEA of Saclay), Centre for Energy Research (CER-EK), Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Helmholtz Zentrum Dresden Rossendorf (HZDR), Institute of Nuclear Technology and Energy Systems (IKE), Japan Atomic Energy Agency (JAEA), Karlsruhe Institute of Technology (KIT

  11. Numerical calculation and analysis of natural convection removal of the spent fuel residual heat of 10 MW high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Wang Jinhua; Huang Yifan; Wu Bin

    2013-01-01

    The spent fuel of 10 MW High Temperature Gas Cooled Reactor (HTR-10) could be stored in the shielded tank, and the tank is stored in the concrete shielded canister in spent fuel storage room, the residual heat of the spent fuel could be removed by the air. The ability of residual heat removal is analyzed in the paper, and the temperature field is numerically calculated through FEA program ANSYS, the analysis and the calculation are used to validate the safety of the spent fuel and the tank, the ultimate temperature of the spent fuel and the tank should below the safety limit. The calculation shows that the maximum temperature locates in the middle of the fuel pebble bed in the spent fuel tank, and the temperature decreases gradually with radial distance, the temperature in the tank body is evenly distributed, and the temperature in the concrete shielded canister decreases gradually with radial distance. It is feasible to remove the residual heat of the spent fuel storage tank by natural ventilation, in natural ventilation condition, the temperature of the spent fuel and the tank is lower than the temperature limit, which provides theoretical evidence for the choice of the residual heat removal method. (authors)

  12. Dynamic Thermal Loads and Cooling Requirements Calculations for V ACs System in Nuclear Fuel Processing Facilities Using Computer Aided Energy Conservation Models

    International Nuclear Information System (INIS)

    EL Fawal, M.M.; Gadalla, A.A.; Taher, B.M.

    2010-01-01

    In terms of nuclear safety, the most important function of ventilation air conditioning (VAC) systems is to maintain safe ambient conditions for components and structures important to safety inside the nuclear facility and to maintain appropriate working conditions for the plant's operating and maintenance staff. As a part of a study aimed to evaluate the performance of VAC system of the nuclear fuel cycle facility (NFCF) a computer model was developed and verified to evaluate the thermal loads and cooling requirements for different zones of fuel processing facility. The program is based on transfer function method (TFM) and it is used to calculate the dynamic heat gain by various multilayer walls constructions and windows hour by hour at any orientation of the building. The developed model was verified by comparing the obtained calculated results of the solar heat gain by a given building with the corresponding calculated values using finite difference method (FDM) and total equivalent temperature different method (TETD). As an example the developed program is used to calculate the cooling loads of the different zones of a typical nuclear fuel facility the results showed that the cooling capacities of the different cooling units of each zone of the facility meet the design requirements according to safety regulations in nuclear facilities.

  13. Investigation of the thermal performance of a vertical two-phase closed thermosyphon as a passive cooling system for a nuclear reactor spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Kusuma, Mukhsinun Hadi; Putra, Nandy; Imawan, Ficky Augusta [Heat Transfer Laboratory, Department of Mechanical Engineering Universitas Indonesia, Kampus (Indonesia); Antariksawan, Anhar Riza [Centre for Nuclear Reactor Safety and Technology, National Nuclear Energy Agency of Indonesia (BATAN), Kawasan Puspiptek Serpong (Indonesia)

    2017-04-15

    The decay heat that is produced by nuclear reactor spent fuel must be cooled in a spent fuel storage pool. A wickless heat pipe or a vertical two-phase closed thermosyphon (TPCT) is used to remove this decay heat. The objective of this research is to investigate the thermal performance of a prototype model for a large-scale vertical TPCT as a passive cooling system for a nuclear research reactor spent fuel storage pool. An experimental investigation and numerical simulation using RELAP5/MOD 3.2 were used to investigate the TPCT thermal performance. The effects of the initial pressure, filling ratio, and heat load were analyzed. Demineralized water was used as the TPCT working fluid. The cooled water was circulated in the water jacket as a cooling system. The experimental results show that the best thermal performance was obtained at a thermal resistance of 0.22°C/W, the lowest initial pressure, a filling ratio of 60%, and a high evaporator heat load. The simulation model that was experimentally validated showed a pattern and trend line similar to those of the experiment and can be used to predict the heat transfer phenomena of TPCT with varying inputs.

  14. Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, Digby; Urquidi-Macdonald, Mirna; Chen, Yingzi; Ai, Jiahe; Park, Pilyeon; Kim, Han-Sang

    2006-12-12

    Our work involved the continued development of the theory of passivity and passivity breakdown, in the form of the Point Defect Model, with emphasis on zirconium and zirconium alloys in reactor coolant environments, the measurement of critically-important parameters, and the development of a code that can be used by reactor operators to actively manage the accumulation of corrosion damage to the fuel cladding and other components in the heat transport circuits in both BWRs and PWRs. In addition, the modified boiling crevice model has been further developed to describe the accumulation of solutes in porous deposits (CRUD) on fuel under boiling (BWRs) and nucleate boiling (PWRs) conditions, in order to accurately describe the environment that is contact with the Zircaloy cladding. In the current report, we have derived expressions for the total steady-state current density and the partial anodic and cathodic current densities to establish a deterministic basis for describing Zircaloy oxidation. The models are “deterministic” because the relevant natural laws are satisfied explicitly, most importantly the conversation of mass and charge and the equivalence of mass and charge (Faraday’s law). Cathodic reactions (oxygen reduction and hydrogen evolution) are also included in the models, because there is evidence that they control the rate of the overall passive film formation process. Under open circuit conditions, the cathodic reactions, which must occur at the same rate as the zirconium oxidation reaction, are instrumental in determining the corrosion potential and hence the thickness of the barrier and outer layers of the passive film. Controlled hydrodynamic methods have been used to measure important parameters in the modified Point Defect Model (PDM), which is now being used to describe the growth and breakdown of the passive film on zirconium and on Zircaloy fuel sheathing in BWRs and PWRs coolant environments. The modified PDMs recognize the existence of a

  15. Hardware/Software Data Acquisition System for Real Time Cell Temperature Monitoring in Air-Cooled Polymer Electrolyte Fuel Cells.

    Science.gov (United States)

    Segura, Francisca; Bartolucci, Veronica; Andújar, José Manuel

    2017-07-09

    This work presents a hardware/software data acquisition system developed for monitoring the temperature in real time of the cells in Air-Cooled Polymer Electrolyte Fuel Cells (AC-PEFC). These fuel cells are of great interest because they can carry out, in a single operation, the processes of oxidation and refrigeration. This allows reduction of weight, volume, cost and complexity of the control system in the AC-PEFC. In this type of PEFC (and in general in any PEFC), the reliable monitoring of temperature along the entire surface of the stack is fundamental, since a suitable temperature and a regular distribution thereof, are key for a better performance of the stack and a longer lifetime under the best operating conditions. The developed data acquisition (DAQ) system can perform non-intrusive temperature measurements of each individual cell of an AC-PEFC stack of any power (from watts to kilowatts). The stack power is related to the temperature gradient; i.e., a higher power corresponds to a higher stack surface, and consequently higher temperature difference between the coldest and the hottest point. The developed DAQ system has been implemented with the low-cost open-source platform Arduino, and it is completed with a modular virtual instrument that has been developed using NI LabVIEW. Temperature vs time evolution of all the cells of an AC-PEFC both together and individually can be registered and supervised. The paper explains comprehensively the developed DAQ system together with experimental results that demonstrate the suitability of the system.

  16. Hardware/Software Data Acquisition System for Real Time Cell Temperature Monitoring in Air-Cooled Polymer Electrolyte Fuel Cells

    Directory of Open Access Journals (Sweden)

    Francisca Segura

    2017-07-01

    Full Text Available This work presents a hardware/software data acquisition system developed for monitoring the temperature in real time of the cells in Air-Cooled Polymer Electrolyte Fuel Cells (AC-PEFC. These fuel cells are of great interest because they can carry out, in a single operation, the processes of oxidation and refrigeration. This allows reduction of weight, volume, cost and complexity of the control system in the AC-PEFC. In this type of PEFC (and in general in any PEFC, the reliable monitoring of temperature along the entire surface of the stack is fundamental, since a suitable temperature and a regular distribution thereof, are key for a better performance of the stack and a longer lifetime under the best operating conditions. The developed data acquisition (DAQ system can perform non-intrusive temperature measurements of each individual cell of an AC-PEFC stack of any power (from watts to kilowatts. The stack power is related to the temperature gradient; i.e., a higher power corresponds to a higher stack surface, and consequently higher temperature difference between the coldest and the hottest point. The developed DAQ system has been implemented with the low-cost open-source platform Arduino, and it is completed with a modular virtual instrument that has been developed using NI LabVIEW. Temperature vs time evolution of all the cells of an AC-PEFC both together and individually can be registered and supervised. The paper explains comprehensively the developed DAQ system together with experimental results that demonstrate the suitability of the system.

  17. Investigation on thermo-acoustic instability dynamic characteristics of hydrocarbon fuel flowing in scramjet cooling channel based on wavelet entropy method

    Science.gov (United States)

    Zan, Hao; Li, Haowei; Jiang, Yuguang; Wu, Meng; Zhou, Weixing; Bao, Wen

    2018-06-01

    As part of our efforts to find ways and means to further improve the regenerative cooling technology in scramjet, the experiments of thermo-acoustic instability dynamic characteristics of hydrocarbon fuel flowing have been conducted in horizontal circular tubes at different conditions. The experimental results indicate that there is a developing process from thermo-acoustic stability to instability. In order to have a deep understanding on the developing process of thermo-acoustic instability, the method of Multi-scale Shannon Wavelet Entropy (MSWE) based on Wavelet Transform Correlation Filter (WTCF) and Multi-Scale Shannon Entropy (MSE) is adopted in this paper. The results demonstrate that the developing process of thermo-acoustic instability from noise and weak signals is well detected by MSWE method and the differences among the stability, the developing process and the instability can be identified. These properties render the method particularly powerful for warning thermo-acoustic instability of hydrocarbon fuel flowing in scramjet cooling channels. The mass flow rate and the inlet pressure will make an influence on the developing process of the thermo-acoustic instability. The investigation on thermo-acoustic instability dynamic characteristics at supercritical pressure based on wavelet entropy method offers guidance on the control of scramjet fuel supply, which can secure stable fuel flowing in regenerative cooling system.

  18. A comparison of fuel savings in the residential and commercial sectors generated by the installation of solar heating and cooling systems under three tax credit scenarios

    Science.gov (United States)

    Moden, R.

    An analysis of expected energy savings between 1977 and 1980 under three different solar tax credit scenarios is presented. The results were obtained through the solar heating and cooling of buildings (SHACOB) commercialization model. This simulation provides projected savings of conventional fuels through the installation of solar heating and cooling systems on buildings in the residential and commercial sectors. The three scenarios analyzed considered the tax credits contained in the Windfall Profits Tax of April 1980, the National Tax Act of November 1978, and a case where no tax credit is in effect.

  19. Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

    1976-04-01

    Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

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

    International Nuclear Information System (INIS)

    Lee, D.

    1995-02-01

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

  1. The effect of porosity on performance of phosphoric acid doped polybenzimidazole polymer electrolyte membrane fuel cell

    Directory of Open Access Journals (Sweden)

    Celik Muhammet

    2016-01-01

    Full Text Available A polybenzimidazole (PBI based polymer electrolyte fuel cells, which called high temperature polymer electrolyte fuel cells (HT-PEMS, operate at higher temperatures (120-200°C than conventional PEM fuel cells. Although it is known that HT-PEMS have some of the significant advantages as non-humidification requirements for membrane and the lack of liquid water at high temperature in the fuel cell, the generated water as a result of oxygen reduction reaction causes in the degradation of these systems. The generated water absorbed into membrane side interacts with the hydrophilic PBI matrix and it can cause swelling of membrane, so water transport mechanism in a membrane electrode assembly (MEA needs to be well understood and water balance must be calculated in MEA. Therefore, the water diffusion transport across the electrolyte should be determined. In this study, various porosity values of gas diffusion layers are considered in order to investigate the effects of porosity on the water management for two phase flow in fuel cell. Two-dimensional fuel cell with interdigitated flow-field is modelled using COMSOL Multiphysics 4.2a software. The operating temperature and doping level is selected as 160°C and 6.75mol H3PO4/PBI, respectively.

  2. Effects of duct configuration on flow and temperature structure in sodium-cooled 19-rod simulated LMFBR fuel bundles with helical wire-wrap spacers

    International Nuclear Information System (INIS)

    Wantland, J.L.; Fontana, M.H.; Gnadt, P.A.; Hanus, N.; MacPherson, R.E.; Smith, C.M.

    1976-01-01

    Thermal-hydrodynamic testing of sodium-cooled 19-rod simulated LMFBR fuel bundles is being conducted at the O ak Ridge National Laboratory in the Fuel Failure Mockup (FFM), an engineering-scale high-temperature sodium facility which provides prototypic flows, temperatures and power densities. Electrically heated bundles have been tested with two scalloped and two hexagonal duct configurations. Peripheral helical flows, attributed to the spacers, have been observed with strengths dependent upon the evenness and relative sizes of the peripheral flow areas. Diametral sodium temperature profiles are more uniform with smaller peripheral flow areas

  3. An investigation on technical feasibilities of fuel cycle for high temperature gas-cooled reactor (Case study)

    International Nuclear Information System (INIS)

    Sumita, Junya; Ueta, Shohei; Aihara, Jun; Shibata, Taiju; Sawa, Kazuhiro

    2008-03-01

    In accordance with the basic policy of effectively using nuclear fuel resources, the FBR cycle, one of the most possible fuel cycle in the future, will be adapted after plu-thermal program by LWR in Japanese nuclear cycle plan. In this paper, a case study of technical investigation of HTGR fuel cycle based on HTGR fuel cycle proposed to adapt to Japanese nuclear fuel cycle plan were carried out from the viewpoint of effective utilization of uranium, fabrication technologies of MOX fuel, reprocessing technologies, amount of interim storage of HTGR fuel and graphite waste. As a result, the fuel cycle for HTGR is expected to be possible technically. (author)

  4. Safety performance comparation of MOX, nitride and metallic fuel based 25-100 MWe Pb-Bi cooled long life fast reactors without on-site refuelling

    International Nuclear Information System (INIS)

    Su'ud, Zaki

    2008-01-01

    In this paper the safety performance of 25-100 MWe Pb-Bi cooled long life fast reactors based on three types of fuels: MOX, nitride and metal is compared and discussed. In the fourth generation NPP paradigm, especially for Pb-Bi cooled fast reactors, inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Selection of fuel type will have important impact on the overall system safety performance. The results of safety analysis of long life Pb-Bi cooled fast reactors without on-site fuelling using nitride, MOX and metal fuel have been performed. The reactors show the inherent safety pattern with enough safety margins during ULOF and UTOP accidents. For MOX fuelled reactors, ULOF accident is more severe than UTOP accident while for nitride fuelled cores UTOP accident may push power much higher than that comparable MOX fuelled cores. (author)

  5. Comparison of MCNPX-C90 and TRIPOLI-4-D for fuel depletion calculations of a Gas-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Reyes-Ramirez, Ricardo; Martin-del-Campo, Cecilia; Francois, Juan-Luis; Brun, Emeric; Dumonteil, Eric; Malvagi, Fausto

    2010-01-01

    The Gas-cooled Fast Reactor is one of the reactor concepts selected by the Generation IV International Forum for the next generation of innovative nuclear energy systems. Several fuel design concepts are being investigated. Burnup depletion of mixed fuel of uranium and plutonium, cooled with gas in a fast neutron energy spectrum must be simulated. Various codes are being developed and/or adapted to improve the quality of the results, and also to reduce the computing time required for the simulations. The main objective of this work is to compare the fuel depletion results obtained with MCNPX-CINDER90 code and the new TRIPOLI-4-Depletion code (developed by the Commissariat a l'Energie Atomique) of a fuel design concept for the Gas-cooled Fast Reactor. Calculations were made for an equivalent homogeneous model of fuel rods in a hexagonal mesh assembly. Total reflection conditions were applied on the six lateral faces and the two axial faces of the assembly. The materials used in the fuel assembly are: carbide of uranium and plutonium as fuel, silicon carbide as cladding, and helium gas as coolant. JEFF libraries of effective cross sections were used in both codes. Two methods of burnup step calculations were performed with TRIPOLI-4-D, the Euler and the CSADA, and their results were compared with the MCNPX-CINDER90 CSADA method. A period of 300 days of irradiation time was considered, which was divided into 12 steps. Results of the infinite multiplication factor as function of the irradiation time, and the evolution of the isotope concentrations for a selected group of nuclides were compared. The main conclusion is that very similar results were obtained for the three types of depletion calculations which were compared: (1) MCNPX-C90 CSADA; (2) TRIPOLI-4-D CSADA, and (3) TRIPOLI-4-D EULER. The best calculation time was obtained with the TRIPOLI-4-D EULER method, which needed approximately half the time than the other two. In summary, it is sufficiently good to use

  6. Modeling and Thermal Performance Evaluation of Porous Curd Layers in Sub-Cooled Boiling Region of PWRs and Effects of Sub-Cooled Nucleate Boiling on Anomalous Porous Crud Deposition on Fuel Pin Surfaces

    International Nuclear Information System (INIS)

    Barclay Jones

    2005-01-01

    A significant number of current PWRs around the world are experiencing anomalous crud deposition in the sub-cooled region of the core, resulting in an axial power shift or Axial Offset Anomaly (AOA), a condition that continues to elude prediction of occurrence and thermal/neutronic performance. This creates an operational difficulty of not being able to accurately determine power safety margin. In some cases this condition has required power ''down rating'' by as much as thirty percent and the concomitant considerable loss of revenue for the utility. This study examines two aspects of the issue: thermal performance of crud layer and effect of sub-cooled nucleate boiling on the solute concentration and its influence on initiation of crud deposition/formation on fuel pin surface

  7. Pressure loss coefficient and flow rate of side hole in a lower end plug for dual-cooled annular nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang-Hwan, E-mail: shinch@kaeri.re.kr; Park, Ju-Yong, E-mail: juyong@kaeri.re.kr; In, Wang-Kee, E-mail: wkin@kaeri.re.kr

    2013-12-15

    Highlights: • A lower end plug with side flow holes is suggested to provide alternative flow paths of the inner channel. • The inlet loss coefficient of the lower end plug is estimated from the experiment. • The flow rate through the side holes is estimated in a complete entrance blockage of inner channel. • The consequence in the reactor core condition is evaluated with a subchannel analysis code. - Abstract: Dual-cooled annular nuclear fuel for a pressurized water reactor (PWR) has been introduced for a significant increase in reactor power. KAERI has been developing a dual-cooled annular fuel for a power uprate of 20% in an optimized PWR in Korea, the OPR1000. This annular fuel can help decrease the fuel temperature substantially relative to conventional cylindrical fuel at a power uprate. Annular fuel has dual flow channels around itself; however, the inner flow channel has a weakness in that it is isolated unlike the outer flow channel, which is open to other neighbouring outer channels for a coolant exchange in the reactor core. If the entrance of the inner channel is, as a hypothetical event, completely blocked by debris, the inner channel will then experience a rapid increase in coolant temperature such that a departure from nucleate boiling (DNB) may occur. Therefore, a remedy to avoid such a postulated accident is indispensable for the safety of annular fuel. A lower end plug with side flow holes was suggested to provide alternative flow paths in addition to the central entrance of the inner channel. In this paper, the inlet loss coefficient of the lower end plug and the flow rate through the side holes were estimated from the experimental results even in a complete entrance blockage of the inner channel. An optimization for the side hole was also performed, and the results are applied to a subchannel analysis to evaluate the consequence in the reactor core condition.

  8. Characterization of velocity and temperature fields in a 217 pin wire wrapped fuel bundle of sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Naveen Raj, M.; Velusamy, K.

    2016-01-01

    Highlights: • We simulate flow and temperature fields in fuel subassembly of fast reactor. • We perform high fidelity computations for 217 pin bundle of 7 axial pitch lengths. • We investigate transverse and axial flows in different types of subchannels. • Correlations are proposed for transverse flow, which form input for subchannel analysis. • Periodic variations of large magnitude are observed in subchannel flow rates. - Abstract: RANS based computational fluid dynamic (CFD) simulation of flow and temperature fields in a fast reactor fuel subassembly has been carried out. The sodium cooled prototype subassembly consists of 217 pins with helical wire spacers. An axial length of seven helical wire pitches has been considered for the study adopting a structured mesh having 36 million points and 84 processors in parallel. The computational model has been validated against in-house and published experimental data for friction factor and Nusselt number. Also, the transverse flow in the central subchannel and swirl flow in the peripheral subchannel are compared against reported experimental data and those computed by subchannel models. The focus of the study is investigation of transverse and axial flows in different types of subchannels. Based on the 3-dimensional CFD study, correlations have been proposed for calculation of transverse flow, which forms an important input for development of subchannel analysis codes. Periodic variations have been observed in the subchannel axial flow rates. For the subchannels located in the central region, the peak to peak variation in the axial flow rate is ∼21% and it is found to be contributed by the changes in the flow area and hydraulic resistance due to frequent passage of helical wires through the subchannel. For the subchannels located in the periphery, this variation is as high as 50%. The transverse flow in the central subchannels follows a cosine profile, for all the faces. However, there is a phase lag of 120

  9. Analysis of Accident Scenarios for the Development of Probabilistic Safety Assessment Model for the Metallic Fuel Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Kim, Tae Woon; Park, S. Y.; Yang, J. E.; Kwon, Y. M.; Jeong, H. Y.; Suk, S. D.; Lee, Y. B.

    2009-03-01

    The safety analysis reports which were reported during the development of sodium cooled fast reactors in the foreign countries are reviewed for the establishment of Probabilistic Safety Analysis models for the domestic SFR which are under development. There are lots of differences in the safety characteristics between the mixed oxide (MOX) fuel SFR and metallic fuel SFR. Metallic fuel SFR is under development in Korea while MOX fuel SFR is under development in France, Japan, India and China. Therefore the status on the development of fast reactors in the foreign countries are reviewed at first and then the safety characteristics between the MOX fuel SFR and the metallic fuel SFR are reviewed. The core damage can be defined as coolant voiding, fuel melting, cladding damage. The melting points of metallic fuel and the MOX fuel is about 1000 .deg. C and 2300 .deg. C, respectively. The high energy stored in the MOX fuel have higher potential to voiding of coolant compared to the possibility in the metallic fuel. The metallic fuel has also inherent reactivity feedback characteristic that the metallic fuel SFR can be shutdown safely in the events of transient overpower, loss of flow, and loss of heat sink without scram. The metallic fuel has, however, lower melting point due to the eutectic formation between the uranium in metallic fuel and the ferrite in metallic cladding. It is needed to identify the core damage accident scenarios to develop Level-1 PSA model. SSC-K computer code is used to identify the conditions in which the core damage can occur in the KALIMER-600 SFR. The accident cases which are analyzed are the triple failure accidents such as unprotected transient over power events, loss of flow events, and loss of heat sink events with impaired safety systems or functions. Through the analysis of the triple failure accidents for the KALIMER-600 SFR, it is found that the PSA model developed for the PRISM reactor design can be applied to KALIMER-600. However

  10. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    International Nuclear Information System (INIS)

    Monado, F.; Permana, S.

    2013-01-01

    Full-text: A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8 % HM. From the neutronic point of view, this design is in compliance with good performance. (author)

  11. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    International Nuclear Information System (INIS)

    Monado, Fiber; Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik; Aziz, Ferhat; Sekimoto, Hiroshi

    2014-01-01

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance

  12. Coolant-fuel interaction in Sodium-cooled Fast Reactors: Structural investigations of The Na-An-O (An = U, Np, Pu) systems

    International Nuclear Information System (INIS)

    Smith, A.L.; Raison, P.E.; Bykov, D.M.; Konings, R.J.; Caciuffo, R.; Cheetham, A.K.

    2014-01-01

    Nuclear energy has the potential to provide Europe with a secure and sustainable electricity supply at a competitive price and to make a significant contribution to the reduction of greenhouse gases emissions. The interest for Sodium-cooled-Fast-spectrum Reactors (SFRs), when compared to Pressurized Water Reactors (PWRs), lies in their more efficient management of plutonium and other actinides as well as their ability to use almost all of the energy in the natural uranium versus 1% utilized in thermal spectrum systems. The high fuel efficiency of fast reactors could greatly dampen concerns about fuel supply. But these reactors have also several drawbacks when compared to PWRs (i.e sodium fire, Na reaction with O2 and H2O, interaction of sodium with oxide fuels). Their development at an industrial scale needs therefore an exhaustive safety assessment that comprises both experimental work and development of sophisticated modelling tools able to describe the reactor behaviour in normal or incidental conditions

  13. Fuel technology and performance of non-water cooled reactors. Proceedings of an advisory group meeting held in Vienna, 5-8 December 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-01

    The IAEA Division of Nuclear Fuel Cycle and Waste Management has been closely involved for many years in the collection, analysis and exchange of information relating to the global development of advanced reactor fuel technology and performance. Meetings of experts in this field have been held in 1984 and 1989 and more recently in December 1994 as part of the IAEA`s programme. This publication reviews progress in advanced reactor fuel technology and performance over the past five years, principally related to non-water cooled reactors, namely high temperature gas reactors (HTGRs) and fast reactors (FRs), as well as developments pertaining to thorium fuels and the fuel fabrication technologies. It includes papers from the participants and provides recommendations in key areas where further global co-operation in this field might be usefully initiated or strengthened. The previous two Advisory Group Meetings on Advanced Fuel Technology and Performance, on which separate reports have been published (IAEA-TECDOC-352 (1985) and IAEA-TECDOC-577 (1990)), focused on all types of commercial nuclear reactors. Refs, figs and tabs.

  14. Design and Control of High Temperature PEM Fuel Cell Systems using Methanol Reformers with Air or Liquid Heat Integration

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Kær, Søren Knudsen; Sahlin, Simon Lennart

    2013-01-01

    The present work describes the ongoing development of high temperature PEM fuel cell systems fuelled by steam reformed methanol. Various fuel cell system solutions exist, they mainly differ depending on the desired fuel used. High temperature PEM (HTPEM) fuel cells offer the possibility of using...... methanol is converted to a hydrogen rich gas with CO2 trace amounts of CO, the increased operating temperatures allow the fuel cell to tolerate much higher CO concentrations than Nafion-based membranes. The increased tolerance to CO also enables the use of reformer systems with less hydrogen cleaning steps...... liquid fuels such as methanol, due to the increased robustness of operating at higher temperatures (160-180oC). Using liquid fuels such as methanol removes the high volume demands of compressed hydrogen storages, simplifies refueling, and enables the use of existing fuel distribution systems. The liquid...

  15. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Favalli, A., E-mail: afavalli@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.J.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg)

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute {sup 137}Cs count rate and the {sup 154}Eu/{sup 137}Cs, {sup 134}Cs/{sup 137}Cs, {sup 106}Ru/{sup 137}Cs, and {sup 144}Ce/{sup 137}Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  16. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Science.gov (United States)

    Favalli, A.; Vo, D.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S. J.; Trellue, H.; Vaccaro, S.

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity's behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  17. Cost analysis and economic comparison for alternative fuel cycles in the heavy water cooled canadian reactor (CANDU)

    International Nuclear Information System (INIS)

    Yilmaz, S.

    2000-01-01

    Three main options in a CANDU fuel cycle involve use of: (1) natural uranium (0.711 weight percent U-235) fuel, (2) slightly enriched uranium (1.2 weight percent U-235) fuel, and (3) recovered uranium (0.83 weight percent U-235) fuel from light water reactor spent fuel. ORIGEN-2 computer code was used to identify composition of the spent fuel for each option, including the standard LWR fuel (3.3 weight percent U-235). Uranium and plutonium credit calculations were performed using ORIGEN-2 output. WIMSD-5 computer code was used to determine maximum discharge burnup values for each case. For the 3 cycles selected (natural uranium, slightly enriched uranium, recovered uranium), levelized fuel cycle cost calculations are performed over the reactor lifetime of 40 years, using unit process costs obtained from literature. Components of the fuel cycle costs are U purchase, conversion, enrichment, fabrication, SF storage, SF disposal, and reprocessing where applicable. Cost parameters whose effects on the fuel cycle cost are to be investigated are escalation ratio, discount rate and SF storage time. Cost estimations were carried out using specially developed computer programs. Share of each cost component on the total cost was determined and sensitivity analysis was performed in order to show how a change in a main cost component affects the fuel cycle cost. The main objective of this study has been to find out the most economical option for CANDU fuel cycle by changing unit prices and cost parameters

  18. Cooling towers

    International Nuclear Information System (INIS)

    Korik, L.; Burger, R.

    1992-01-01

    What is the effect of 0.6C (1F) temperature rise across turbines, compressors, or evaporators? Enthalpy charts indicate for every 0.6C (1F) hotter water off the cooling tower will require an additional 2 1/2% more energy cost. Therefore, running 2.2C (4F) warmer due to substandard cooling towers could result in a 10% penalty for overcoming high heads and temperatures. If it costs $1,250,000.00 a year to operate the system, $125,000.00 is the energy penalty for hotter water. This paper investigates extra fuel costs involved in maintaining design electric production with cooling water 0.6C (1F) to 3C (5.5F) hotter than design. If design KWH cannot be maintained, paper will calculate dollar loss of saleable electricity. The presentation will conclude with examining the main causes of deficient cold water production. State-of-the-art upgrading and methodology available to retrofit existing cooling towers to optimize lower cooling water temperatures will be discussed

  19. Performance Degradation Tests of Phosphoric Acid Doped PBI Membrane Based High Temperature PEM Fuel Cells

    DEFF Research Database (Denmark)

    Zhou, Fan; Araya, Samuel Simon; Grigoras, Ionela

    2014-01-01

    Degradation tests of two phosphoric acid (PA) doped PBI membrane based HT-PEM fuel cells were reported in this paper to investigate the effects of start/stop and the presence of methanol in the fuel to the performance degradation. Continuous tests with H2 and simulated reformate which was composed...... of H2, water steam and methanol as the fuel were performed on both single cells. 12-h-startup/12-h-shutdown dynamic tests were performed on the first single cell with pure dry H2 as the fuel and on the second single cell with simulated reformate as the fuel. Along with the tests electrochemical...... techniques such as polarization curves and electrochemical impedance spectroscopy (EIS) were employed to study the degradation mechanisms of the fuel cells. Both single cells showed an increase in the performance in the H2 continuous tests, because of a decrease in the ORR kinetic resistance probably due...

  20. Study on the possibility of supercritical fluid extraction for reprocessing of spent nuclear fuel from high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Duan Wuhua; Zhu Liyang; Zhu Yongjun; Xu Jingming

    2011-01-01

    International interest in high temperature gas-cooled reactor (HTGR) has been increasing in recent years. It is important to study on reprocessing of spent nuclear fuel from HTGR for recovery of nuclear resource and reduction of nuclear waste. Treatment of UO 2 pellets for preparing fuel elements of the 10 MW high temperature gas-cooled reactor (HTR-10) using supercritical fluid extraction was investigated. UO 2 pellets are difficult to be directly dissolved and extracted with TBP-HNO 3 complex in supercritical CO 2 (SC-CO 2 ), and the extraction efficiency is only about 7% under experimental conditions. UO 2 pellets are also difficult to be converted completely into nitrate with N 2 O 4 . When UO 2 pellets break spontaneously into U 3 O 8 powders with particle size below 100 μm under O 2 flow and 600degc, the extraction efficiency of U 3 O 8 powders with TBP-HNO 3 complex in SC-CO 2 can reach more than 98%. U 3 O 8 powders are easy to be completely converted into nitrate with N 2 O 4 . The extraction efficiency of the nitrate product with TBP in SC-CO 2 can reach more than 99%. So it has a potential prospect that application of supercritical fluid extraction in reprocessing of spent nuclear fuel from HTGR. (author)

  1. ComfortPower. Design, construction and evaluation of a combined fuel-cell and heat pump system

    Energy Technology Data Exchange (ETDEWEB)

    Silversand, Fredrik (Catator AB, Lund (Sweden))

    2010-12-15

    Catator AB has constructed, commissioned and evaluated a combined fuel-cell and heat-pump system (ComfortPower). The basic idea behind the project was to demonstrate the possibility to achieve ultrahigh thermal efficiencies when combining fuel-cell technologies and heat pumps. Moreover, the system should provide a great flexibility with respect to the fuel mix and should in addition to heat provide surplus electricity and cooling. The system was built on a HT-PEM platform (high temperature polymer electrolyte fuel cell from Serenergy a/s), which was operated by Catators proprietary Optiformer technology. The power generator was combined with a heat pump module (F1145-5, 230 V), supplied by Nibe. The system was packaged into a cabinet (1.65 x 0.6 x 0.6 m) comprising the power module, the heat pump, all necessary balance-of-plant components and the control system. The power output from the fuel-cell system was around 1.35 kW, which enabled operation of the heat pump compressor. By utilizing surplus heat energy from the fuel cell it was possible to achieve a favourable operation point in the heat pump system, resulting in a high overall COP (coefficient of performance). The heat output from the system was as high as 10 kW whereas 6 kW cooling could be provided. The thermal efficiencies measured in experiments were normally around 200%, calculated on the lower heating value of the fuel. A number of fuels have been investigated in the fuel cell system, including both gaseous (natural gas/LPG) and liquid fuels (alcohols and kerosene). Indeed, the system has a wide fuel flexibility, which opens up for a variety of applications in campus villages and buildings. This study has demonstrated the possibility to reduce the carbon dioxide footprint by a factor of 2 over conventional boilers in heating applications. In addition the unit can be operated on a variety of fuels and can produce cooling and electricity in addition to heat. A fully working system has been designed

  2. Safety analyses for sodium-cooled fast reactors with pelletized and sphere-pac oxide fuels within the FP-7 European project PELGRIMM - 15386

    International Nuclear Information System (INIS)

    Maschek, W.; Andriolo, L.; Matzerath-Boccaccini, C.; Delage, F.; Parisi, C.; Del Nevo, A.; Abbate, G.; Schmitt, D.

    2015-01-01

    The European FP-7 project PELGRIMM addresses the development of Minor-Actinide (MA) bearing oxide fuel for Sodium-cooled Fast Reactors. Optionally, both MA homogeneous recycling and heterogeneous recycling is investigated with pellet and sphere-pac fuel. A first safety assessment of sphere-pac fuelled cores should be given in the Work Package 4 of the project. This assessment is in continuity with the former FP-7 CP-ESFR project. Within the CP-ESFR project the CONF2 core design has been developed characterized by a core with a large upper sodium plenum to reduce the coolant void worth. This optimized core has been chosen for the safety analyses in PELGRIMM. The task within the PELGRIMM project is thus a safety assessment of the CONF2 core loaded either with pellets or with sphere-pac fuel. The investigations started with the design of the CONF2 core with sphere-pac fuel and the determination of core safety parameters and burn-up behavior. The neutronic analyses have been performed with the MCNPX code. Variants of the CONF2 core contain up to 4% Am in the fuel. The results revealed an extended void worth (core + upper plenum) for an Am free core of 1 up to 3 dollars for the 4% Am core. Thermal-hydraulic design analyses have been performed by RELAP5-3D. The accident simulations should be performed by different codes, some of which focus on the initiation phase of the accident, as SAS4A, BELLA and the MAT5DYN code, whereas the SIMMER-III code will also deal with the later accident phases and a potential whole core melting. The codes had to be adapted to the specifics of the sphere-pac fuel, in particular to the thermal conductivity and gap conditions. Analyses showed that the safety assessment has to take into account two main phases. Starting up the core, the green fuel shows a reduced fuel thermal conductivity. After restructuring within a couple of hours, the thermal conductivity recovers and the fuel temperature decreases. The main objective of the safety analyses

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

    International Nuclear Information System (INIS)

    Chen, S.R.; Lin, W.C.; Ferng, Y.M.; Chieng, C.C.; Pei, B.S.

    2014-01-01

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

  4. Neutronics optimization of LiPb-He dual-cooled fuel breeding blanket for the fusion-driven sub-critical system

    International Nuclear Information System (INIS)

    Zheng Shanliang; Wu Yican

    2002-01-01

    The concept of the liquid Li 17 Pb 83 and Helium gas dual-cooled Fuel Breeding Blanket (FBB) for the Fusion-Driven sub-critical System (FDS) is presented and analyzed. Taking self-sustaining tritium (TBR > 1.05) and annual output of 100 kg or more fissile 239 Pu (FBR > 0.238) as objective parameters, and based on the three-dimensional Monte Carlo neutron-photon transport code MCNP/4A, a neutronics-optimized calculation of different cases was carried out and the concept is proved feasible. In addition, the total breeding ratio (Br = Tbr + Fbr) is listed corresponding to different cases

  5. Process and device for finding and alarming faults in cooling in a fuel element of a reactor core

    International Nuclear Information System (INIS)

    Holick, A.

    1985-01-01

    The coolant outlet temperature of each fuel element of a BWR is no longer monitored, but the mean fuel or coolant temperature, which cannot be measured directly, is derived from parameters which can easily be measured using a thermo-hydraulic model. This mean fuel or coolant temperature is a parameter, whose value is only affected by very little noise. The problems which arose from the noise of the coolant outlet temperature no longer occur when determining the mean fuel temperature. A deviation of the mean fuel or coolant temperature from the reference value can therefore be found with great accuracy, so that a fault alarm can be initiated for the fuel element concerned. (orig./HP) [de

  6. Investigation of Reactivity Feedback Mechanism of Axial and Radial Expansion Effect of Metal-Fueled Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Seong, Seung-Hwan; Choi, Chi-Woong; Jeong, Tae-Kyung; Ha, Gi-Seok

    2015-01-01

    The major inherent reactivity feedback models for a ceramic fuel used in a conventional light water reactor are Doppler feedback and moderator feedback. The metal fuel has these two reactivity feedback mechanisms previously mentioned. In addition, the metal fuel has two more reactivity feedback models related to the thermal expansion phenomena of the metal fuel. Since the metal fuel has a good capability to expand according to the temperature changes of the core, two more feedback mechanisms exist. These additional two feedback mechanism are important to the inherent safety of metal fuel and can make metal-fueled SFR safer than oxide-fueled SFR. These phenomena have already been applied to safety analysis on design extended condition. In this study, the effect of these characteristics on power control capability was examined through a simple load change operation. The axial expansion mechanism is induced from the change of the fuel temperature according to the change of the power level of PGSFR. When the power increases, the fuel temperatures in the metal fuel will increase and then the reactivity will decrease due to the axial elongation of the metal fuel. To evaluate the expansion effect, 2 cases were simulated with the same scenario by using MMS-LMR code developed at KAERI. The first simulation was to analyze the change of the reactor power according to the change of BOP power without the reactivity feedback model of the axial and radial expansion of the core during the power transient event. That is to say, the core had only two reactivity feedback mechanism of Doppler and coolant temperature

  7. Review of experimental studies of zirconium carbide coated fuel particles for high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Minato, Kazuo; Ogawa, Toru; Fukuda, Kousaku

    1995-03-01

    Experimental studies of zirconium carbide(ZrC) coated fuel particles were reviewed from the viewpoints of fuel particle designs, fabrication, characterization, fuel performance, and fission product retentiveness. ZrC is known as a refractory and chemically stable compound, so ZrC is a candidate to replace the silicon carbide(SiC) coating layer of the Triso-coated fuel particles. The irradiation experiments, the postirradiation heating tests, and the out-of-reactor experiments showed that the ZrC layer was less susceptible than the SiC layer to chemical attack by fission products and fuel kernels, and that the ZrC-coated fuel particles performed better than the standard Triso-coated fuel particles at high temperatures, especially above 1600degC. The ZrC-coated fuel particles demonstrated better cesium retention than the standard Triso-coated fuel particles though the ZrC layer showed a less effective barrier to ruthenium than the SiC layer. (author) 51 refs

  8. Development of variable-width ribbon heating elements for liquid-metal and gas-cooled fast breeder reactor fuel-pin simulators

    International Nuclear Information System (INIS)

    McCulloch, R.W.; Post, D.W.; Lovell, R.T.; Snyder, S.D.

    1981-04-01

    Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relate this profile to that generated by the coils in completed fuel pin simulators

  9. Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

    2014-12-15

    Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

  10. Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Maldonado, Ivan [Univ. of Tennessee, Knoxville, TN (United States)

    2016-04-14

    The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate (“plank”) fuel. Proposal to FY12 NEUP entitled “Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors” was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed on December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project’s success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.

  11. Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

    International Nuclear Information System (INIS)

    Petrovic, Bojan; Maldonado, Ivan

    2016-01-01

    The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate ('plank') fuel. Proposal to FY12 NEUP entitled 'Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors' was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed on December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project's success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.

  12. Numerical evaluation of various gas and coolant channel designs for high performance liquid-cooled proton exchange membrane fuel cell stacks

    International Nuclear Information System (INIS)

    Sasmito, Agus P.; Kurnia, Jundika C.; Mujumdar, Arun S.

    2012-01-01

    A careful design of gas and coolant channel is essential to ensure high performance and durability of proton exchange membrane (PEM) fuel cell stack. The channel design should allow for good thermal, water and gas management whilst keeping low pressure drop. This study evaluates numerically the performance of various gas and coolant channel designs simultaneously, e.g. parallel, serpentine, oblique-fins, coiled, parallel-serpentine and a novel hybrid parallel-serpentine-oblique-fins designs. The stack performance and local distributions of key parameters are investigated with regards to the thermal, water and gas management. The results indicate that the novel hybrid channel design yields the best performance as it constitutes to a lower pumping power and good thermal, water and gas management as compared to conventional channels. Advantages and limitation of the designs are discussed in the light of present numerical results. Finally, potential application and further improvement of the design are highlighted. -- Highlights: ► We evaluate various gas and coolant channel designs in liquid-cooled PEM fuel cell stack. ► The model considers coupled electrochemistry, channel design and cooling effect simultaneously. ► We propose a novel hybrid channel design. ► The novel hybrid channel design yields the best thermal, water and gas management which is beneficial for long term durability. ► The novel hybrid channel design exhibits the best performance.

  13. Minimum Specific Fuel Consumption of a Liquid-Cooled Multicylinder Aircraft Engine as Affected by Compression Ratio and Engine Operating Conditions

    Science.gov (United States)

    Brun, Rinaldo J.; Feder, Melvin S.; Harries, Myron L.

    1947-01-01

    An investigation was conducted on a 12-cylinder V-type liquid-cooled aircraft engine of 1710-cubic-inch displacement to determine the minimum specific fuel consumption at constant cruising engine speed and compression ratios of 6.65, 7.93, and 9.68. At each compression ratio, the effect.of the following variables was investigated at manifold pressures of 28, 34, 40, and 50 inches of mercury absolute: temperature of the inlet-air to the auxiliary-stage supercharger, fuel-air ratio, and spark advance. Standard sea-level atmospheric pressure was maintained at the auxiliary-stage supercharger inlet and the exhaust pressure was atmospheric. Advancing the spark timing from 34 deg and 28 deg B.T.C. (exhaust and intake, respectively) to 42 deg and 36 deg B.T.C. at a compression ratio of 6.65 resulted in a decrease of approximately 3 percent in brake specific fuel consumption. Further decreases in brake specific fuel consumption of 10.5 to 14.1 percent (depending on power level) were observed as the compression ratio was increased from 6.65 to 9.68, maintaining at each compression ratio the spark advance required for maximum torque at a fuel-air ratio of 0.06. This increase in compression ratio with a power output of 0.585 horsepower per cubic inch required a change from . a fuel- lend of 6-percent triptane with 94-percent 68--R fuel at a compression ratio of 6.65 to a fuel blend of 58-percent, triptane with 42-percent 28-R fuel at a compression ratio of 9.68 to provide for knock-free engine operation. As an aid in the evaluation of engine mechanical endurance, peak cylinder pressures were measured on a single-cylinder engine at several operating conditions. Peak cylinder pressures of 1900 pounds per square inch can be expected at a compression ratio of 9.68 and an indicated mean effective pressure of 320 pounds per square inch. The engine durability was considerably reduced at these conditions.

  14. Problems of creating fuel elements for fast gas-cooled reactors working on N2O4-dissociating coolant

    International Nuclear Information System (INIS)

    Nesterenko, V.B.; Zelensky, V.F.; Kolykhan, L.I.; Karpenko, G.V.; Krasnorutsky, V.S.; Isakov, V.P.; Ashikhmin, V.P.; Permyakov, L.N.

    1985-01-01

    A variant of fast gas-cooled reactors is one using dissociating N 2 O 4 nitrogen tetroxide as a coolant. This type of reactors is promising because of great thermal effects of dissociation reactions while heating and recombination while cooling; small latent heat of evaporation; high heat transfer coefficient owing to additional heat transfer in a chemical reaction; high N 2 O 4 density in a gas state at operation parameters. The mentioned advantages give possibility to create a small turbine, heat exchange apparatus and to get high heat production in the active zone. All this opens new ways to increase power plants effectiveness

  15. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

  16. Development and application of an advanced fuel model for the safety analysis of the generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Petkevich, P.

    2008-09-01

    Until about the year 2030, current-day nuclear power plants (NPPs) will be replaced by so-called Gen-III or Gen-III+ units, which are mainly based on light water reactor technology. The principal new features are increased safety and improved economical effectiveness. However, these systems use the same fuel forms and are based on the same fuel cycle. Beyond 2030, the interest is likely to shift towards fourth generation NPPs, which offer the possibility of complete fuel cycle closure. Generation-IV reactor concepts include both thermal and fast systems, and involve a wide range of fuel forms and compositions. The present research has been focused on the development of a thermo-mechanical model for the innovative fuel design of the Generation-IV Gas-cooled Fast Reactor (GFR). The principal distinctive feature of the fuel is that the fuel pellets are arranged within plates which enclose an inner honeycomb structure. Apart from the geometry, the usage of new materials is foreseen. Thus, the fuel pellets are of mixed uranium-plutonium carbide, and the cladding is bulk or fiber-reinforced SiC. The setting up of an appropriate materials database was thus the very first task which had to be carried out in the current work. The main purpose of the currently developed model is to provide reliable data, in the context of transient analysis, for the calculation of the principal neutronic feedbacks in the GFR core, viz. the fuel temperature for the Doppler effect and the fuel plate deformation for the axial core expansion effect. None of the available fuel modeling codes is suitable for a realistic simulation of the GFR fuel, as the inner honeycomb structure cannot be explicitly taken into account. The development work has been carried out largely in the context of PSI’s generic code system for fast reactor safety analysis, FAST. Thereby, it has mainly involved extension of the thermo-mechanical code FRED, developed originally for the modeling of traditional rodded fuel

  17. Response of fuel, fuel elements and gas cooled reactor cores under accidental air or water ingress conditions. Proceedings of a technical committee meeting held in Beijing, China, 25-27 October 1993

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-01-01

    The meeting was convened by the IAEA on the recommendation of the IAEA`s International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Netherlands, Switzerland, the Russian Federation, the United Kingdom and the United States of America. The meeting covered the following topics: experimental investigations of the effects of air and water ingress; predicted response of fuel, graphite and other reactor components; options for minimizing or mitigating the effects of air or water ingress. 19 papers were presented at the meeting. A separate abstract was prepared for each of these papers. Refs, figs and tabs.

  18. Detection of fuel element failures at the gas cooled reactors; Zaznavanje razpok na gorilnih elementih plinsko hlajenega reaktorja

    Energy Technology Data Exchange (ETDEWEB)

    Bremsak, P; Ranzel, S; Ogrin, R; Pisanski, C [Institut Jozef Stefan, Ljubljana (Yugoslavia)

    1964-07-01

    This paper describes the system for measuring the concentration of released Kr and Xe in the cooling gas. The system developed in the Jozef Stefan Institute is efficient and suitable for application The method is based on electrostatic collection of daughter elements from radioactive decay of Xe and Kr.

  19. Experimental determination of the heat transfer coefficient for the optimal design of the cooling system of a PEM fuel cell placed inside the fuselage of an UAV

    International Nuclear Information System (INIS)

    Barroso, Jorge; Renau, Jordi; Lozano, Antonio; Miralles, José; Martín, Jesús; Sánchez, Fernando; Barreras, Félix

    2015-01-01

    The objective of this research is to calculate the heat transfer coefficients needed for the further design of the optimal cooling system of a high-temperature polymer electrolyte membrane fuel cell (HT-PEMFC) stack that will be incorporated to the powerplant of a light unmanned aerial vehicle (UAV) capable of reaching an altitude of 10,000 m. Experiments are performed in two rectangular tunnels, for three different form factors, in experimental conditions as close as possible to the actual ones in the HT-PEMFC stack. For the calculations, all the relevant thermal processes are considered (i.e., convection and radiation). Different parameters are measured, such as air mass flow rate, inlet and outlet air temperatures, and wall temperatures for bipolar plates and endplates. Different numerical models are fitted revealing the influence of the diverse relevant non-dimensional groups on the Nusselt number. Heat transfer coefficients calculated for the air cooling flow vary from 8 to 44 W m"−"2 K"−"1. Results obtained at sea level are extrapolated for a flight ceiling of 10 km. The flow section is optimized as a function of the power required to cool the stack down to the temperature recommended by the membrane-electrode assembly (MEA) manufacturer using a numerical code specifically developed for this purpose. - Highlights: • Heat transfer coefficients to refrigerate a HT-PEMFC stack are calculated. • Experiments are performed in 2 wind tunnels, for 3 form factors and real conditions. • The calculated heat transfer coefficient varies from 8 to 44 W m"−"2 K"−"1. • Results at sea level are suitably extrapolated for a target altitude of 10 km. • Flow area is optimized as a function of the power required to cool the stack down.

  20. Fuel and core design study of the sodium-cooled fast reactors. Studies on metallic fuel cores in the JFY2002

    International Nuclear Information System (INIS)

    Sugino, Kazuteru; Mizuno, Tomoyasu

    2003-06-01

    Based on the results obtained in the former feasibility study, the metallic fueled core of ordinary-type, that is, 2-region homogeneous core, has been established aiming at the improvement in the core performance, and subsequent comparison has been performed with the mixed oxide fueled core. Further, the attractive concept of the metallic fueled core of high outlet temperature has been constructed which has good nuclear features as a metallic fueled core and has identical outlet temperature to mixed oxide fuelled core. Following items have been found as a result of the investigation on the ordinary-type core. The metallic fueled core whose maximum fast neutron fluence (En>0.1MeV) is set identical (5x10 23 n/cm 2 ) to the mixed oxide fueled cores with core discharge burnup 150GWd/t has sufficient core performances as a metallic fueled core, e.g. higher breeding ratio and longer operation period compared with mixed oxide fueled cores, but the core discharge burnup is limited up to 100GWd/t. However effective discharge burnup including the contribution of the blanket region is comparative to mixed oxide cores under the same breeding ratio condition. In order to enlarge the core discharge burnup to 150GWd/t keeping the core performance identical to above mentioned core's, the irradiation deformation of structural material should be reduced to that of mixed oxide fueled cores. Further the maximum fast neutron fluence reaches to 7-8x10 23 n/cm 2 (En>0.1MeV). The investigations on the core of high outlet temperature have clarified following items. Even in the change of core regions by pin-diameter form 3-region to 2-region and in the limited maximum fuel pin diameter 8.5 mm, realization of the identical outlet/inlet temperatures to the mixed oxide cores (550/395degC) is feasible under the criteria of the maximum temperature 650degC at the inner surface of the cladding. The constructed core accommodates the targets of breeding ratio from about 1.0 to 1.2 only by adjusting

  1. Comparative study for axial and radial shuffling scheme effect on the performance of Pb-Bi cooled fast reactors with natural uranium as fuel cycle input

    International Nuclear Information System (INIS)

    Zaki Suud; Indah Rosidah; Maryam Afifah; Ferhat Aziz; Sekimoto, H.

    2013-01-01

    Full text:Comparative study for the Design of Pb-Bi cooled fast reactors with natural uranium as fuel cycle input using special radial shuffling strategy and axial direction modified CANDLE burn-up scheme has been performed. The reactors utilizes UN-PuN as fuel, Eutectic Pb-Bi as coolant, and can be operated without refueling for 10 years in each batch. Reactor design optimization is performed to utilize natural uranium as fuel cycle input. This reactor subdivided into 6-10 regions with equal volume in radial directions. The natural uranium is initially put in region 1, and after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions. The calculation has been done by using SRAC-Citation system code and JENDL-3.2 library. The effective multiplication factor change increases monotonously during 10 years reactor operation time. There is significant power distribution change in the central part of the core during the BOC and the EOC in the radial shuffling system. It is larger than that in the case of modified CANDLE case which use axial direction burning region move. The burn-up level of fuel is slowly grows during the first 15 years but then grow faster in the rest of burn-up history. This pattern is a little bit different from the case of modified CANDLE burn-up scheme in Axial direction in which the slow growing burn-up period is relatively longer almost half of the burn-up history. (author)

  2. Current and future research on corrosion and thermalhydraulic issues of HLM cooled reactors and on LMR fuels for fast reactor systems

    International Nuclear Information System (INIS)

    Knebel, J.U.; Konings, R.J.M.

    2002-01-01

    Heavy liquid metals (HLM) such as lead (Pb) or lead-bismuth eutectic (Pb-Bi) are currently investigated world-wide as coolant for nuclear power reactors and for accelerator driven systems (ADS). Besides the advantages of HLM as coolant and spallation material, e.g. high boiling point, low reactivity with water and air and a high neutron yield, some technological issues, such as high corrosion effects in contact with steels and thermalhydraulic characteristics, need further experimental investigations and physical model improvements and validations. The paper describes some typical HLM cooled reactor designs, which are currently considered, and outlines the technological challenges related to corrosion, thermalhydraulic and fuel issues. In the first part of the presentation, the status of presently operated or planned test facilities related to corrosion and thermalhydraulic questions will be discussed. First approaches to solve the corrosion problem will be given. The approach to understand and model thermalhydraulic issues such as heat transfer, turbulence, two-phase flow and instrumentation will be outlined. In the second part of the presentation, an overview will be given of the advanced fuel types that are being considered for future liquid metal reactor (LMR) systems. Advantages and disadvantages will be discussed in relation to fabrication technology and fuel cycle considerations. For the latter, special attention will be given to the partitioning and transmutation potential. Metal, oxide and nitride fuel materials will be discussed in different fuel forms and packings. For both parts of the presentation, an overview of existing co-operations and networks will be given and the needs for future research work will be identified. (authors)

  3. Experimental Characterization and Modeling of PEM Fuel Cells

    DEFF Research Database (Denmark)

    Jespersen, Jesper Lebæk

    fundamental knowledge of the transport and electrochemical processes of PEM fuel cells and to provide methods for obtaining high quality data for PEM fuel cell simulation model validation. In this thesis three different areas of experimental characterization techniques was investigated, they include: Stack...... for obtaining very detailed data of the manifold flow. Moreover, the tools complement each other well, as high quality validation data can be obtained from PIV measurements to verify CFD models. AC Impedance Spectroscopy was used to thoroughly characterize a HTPEM single cell. The measurement method...... was furthermore transferred onto a Labview platform, which signiffcantly improves the exibility and lowers the cost of using this method. This technique is expected to bea very important future tool, used both for material characterization, celldiagnostic, system optimization and as a control input parameter...

  4. Assessment of Proliferation Resistance of Closed Nuclear Fuel Cycle System with Sodium Cooled Fast Reactors Using INPRO Evaluation Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young In; Hahn, Do Hee; Won, Byung Chool; Lee, Dong Uk

    2007-11-15

    Using the INPRO methodology, the proliferation resistance of an innovative nuclear energy system(INS) defined as a closed nuclear fuel cycle system consisting of KALIMER and pyroprocessing, has been assessed. Considering a very early development stage of the INS concept, the PR assessment is carried out based on intrinsic features, if required information and data are not available. The PR assessment of KALIMER and JSFR using the INPRO methodology affirmed that an adequate proliferation resistance has been achieved in both INSs CNFC-SFR, considering the assessor's progress and maturity of design development. KALIMER and JSFR are developed or being developed conforming to the targets and criteria defined for developing Gen IV nuclear reactor system. Based on these assessment results, proliferation resistance and physical protection(PR and PP) of KALIMER and JSFR are evaluated from the viewpoint of requirements for future nuclear fuel cycle system. The envisioned INSs CNFC-SFR rely on active plutonium management based on a closed fuel cycle, in which a fissile material is recycled in an integrated fuel cycle facility within proper safeguards. There is no isolated plutonium in the closed fuel cycle. The material remains continuously in a sequence of highly radioactive matrices within inaccessible facilities. The proliferation resistance assessment should be an ongoing analysis that keeps up with the progress and maturity of the design of Gen IV SFR.

  5. Assessment of Proliferation Resistance of Closed Nuclear Fuel Cycle System with Sodium Cooled Fast Reactors Using INPRO Evaluation Methodology

    International Nuclear Information System (INIS)

    Kim, Young In; Hahn, Do Hee; Won, Byung Chool; Lee, Dong Uk

    2007-11-01

    Using the INPRO methodology, the proliferation resistance of an innovative nuclear energy system(INS) defined as a closed nuclear fuel cycle system consisting of KALIMER and pyroprocessing, has been assessed. Considering a very early development stage of the INS concept, the PR assessment is carried out based on intrinsic features, if required information and data are not available. The PR assessment of KALIMER and JSFR using the INPRO methodology affirmed that an adequate proliferation resistance has been achieved in both INSs CNFC-SFR, considering the assessor's progress and maturity of design development. KALIMER and JSFR are developed or being developed conforming to the targets and criteria defined for developing Gen IV nuclear reactor system. Based on these assessment results, proliferation resistance and physical protection(PR and PP) of KALIMER and JSFR are evaluated from the viewpoint of requirements for future nuclear fuel cycle system. The envisioned INSs CNFC-SFR rely on active plutonium management based on a closed fuel cycle, in which a fissile material is recycled in an integrated fuel cycle facility within proper safeguards. There is no isolated plutonium in the closed fuel cycle. The material remains continuously in a sequence of highly radioactive matrices within inaccessible facilities. The proliferation resistance assessment should be an ongoing analysis that keeps up with the progress and maturity of the design of Gen IV SFR

  6. Test of hybrid power system for electrical vehicles using a lithium-ion battery pack and a reformed methanol fuel cell range extender

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Ashworth, Leanne; Sahlin, Simon Lennart

    2014-01-01

    is delivered by a lithium ion battery pack. In order to increase the run time of the application connected to this battery pack, a high temperature PEM (HTPEM) fuel cell stack acts as an on-board charger able to charge a vehicle during operation as a series hybrid. Because of the high tolerance to carbon...... a down-sized version of the battery pack used in the Mitsubishi iMiEV, which is subjected to power cycles derived from simulations of the vehicle undergoing multiple New European Drive Cycles (NEDC)....

  7. Determination of mixture coefficients in tests simulating channels of a fuel assembly in a water cooled reactor

    International Nuclear Information System (INIS)

    Ferreira, W.R.

    1983-09-01

    Here, the mixture coefficients are determined in tests which simulate two symmetric and two assymetric coolant channels of a water cooled reactor. It was studies the effects of coolant flow and of the distance among the coolant channels on these coefficients. The technique used to determine the mixture coefficients was to introduce a tracer (methylene blue) into the flow. The determination of the tracer concentration at the end of the channels was made by means of spectrophotometric analysis. (author)

  8. Core design studies on various forms of coolants and fuel materials. 2. Studies on liquid heavy metal and gas cooled cores, small cores and evaluation of 4-type cores

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Sakashita, Yoshiyuki; Naganuma, Masayuki; Takaki, Naoyuki; Mizuno, Tomoyasu; Ikegami, Tetsuo

    2001-01-01

    Alternative concepts to sodium cooled fast reactors, such as heavy metal liquid cooled reactors and gas cooled fast reactors were studied in Phase-1 of the feasibility studies, aiming at simplification of the system, high thermal efficiency and enhancing safety. Fuel and core specifications and nuclear characteristics were surveyed to meet the targets for commercialization of fast reactor cycle. Nuclear characteristics of small fast reactor cores were also surveyed from the perspective of the possibility of multi-purpose use and dispersed power stations. The key points of the design study for each concept in Phase-2 were summarized from the aspect of the screening of the candidates for FR commercialization. (author)

  9. Metal hydride store for hydrogen supply and cooling of fuel cell vehicles; Metallhydridspeicher zur Wasserstoffversorgung und Kuehlung von Brennstoffzellenfahrzeugen

    Energy Technology Data Exchange (ETDEWEB)

    Wenger, David

    2009-07-01

    In the context of the author's work, a compact, dynamic metal hydride store was developed which in addition to storing hydrogen can also support the thermomanagement of fuel cell vehicles in extreme situations. The requirements were identified using a semiphysical model of a fuel cell vehicle, and a store was dimensioned accordingly. Additionally, a metal hydride store model was developed on the basis of the balance equations. The model was validated by experiments on a specially designed and constructed store. The simulations enable the optimisation of the store geometry and the prediction of its efficiency in a given operating cycle. (orig.)

  10. Cooling techniques

    International Nuclear Information System (INIS)

    Moeller, S.P.

    1994-01-01

    After an introduction to the general concepts of cooling of charged particle beams, some specific cooling methods are discussed, namely stochastic, electron and laser cooling. The treatment concentrates on the physical ideas of the cooling methods and only very crude derivations of cooling times are given. At the end three other proposed cooling schemes are briefly discussed. (orig.)

  11. Costs and the environmental impact of radioactive waste treatment in reprocessing high-temperature gas-cooled reactor fuel

    International Nuclear Information System (INIS)

    Davis, W. Jr.

    1976-01-01

    A cost-benefit analysis and an analysis of the reduction in population dose from the use of different decontamination equipment in the off-gas system of a model plant for processing spent fuel from HTGR type reactors are presented

  12. High-temperature gas-cooled reactor fuel recycle development. Annual progress report for period ending September 30, 1977

    International Nuclear Information System (INIS)

    Lotts, A.L.; Kasten, P.R.

    1978-09-01

    The status of the following tasks is reported: program management, studies and analysis, fuel processing, refabrication development, in-plant waste treatment, research general support, and major facilities including HTGR recycle reference facility, hot engineering test facility and cold prototype test facility-refabrication

  13. Cool? Young people investigate living in cold housing and fuel poverty. A mixed methods action research study

    Directory of Open Access Journals (Sweden)

    Kimberley C. O'Sullivan

    2017-12-01

    Conclusion: The integrated results confirm that cold housing and risk of fuel poverty are important problems for young people in New Zealand. Results contribute to the evidence-base for policy targeting of schemes such as the Government-sponsored retrofitting of insulation to households with dependent children.

  14. Solar energy as an alternate energy source to mixed oxide fuels in light-water cooled reactors

    International Nuclear Information System (INIS)

    Bertini, H.W.

    1977-01-01

    Supplemental information pertaining to the generic environmental impact statement on the Pu recycling process for mixed oxide light-water cooled reactors (GESMO) was requested from several sources. In particular, the role of alternate sources of energy was to be explored and the implications of these alternate sources to the question of Pu recycle in LWRs were to be investigated. In this vein, solar energy as an alternate source is the main subject of this report, along with other information related to solar energy. The general conclusion is that solar energy should have little effect on the decisions concerning GESMO

  15. CFD study of liquid-cooled heat sinks with microchannel flow field configurations for electronics, fuel cells, and concentrated solar cells

    International Nuclear Information System (INIS)

    Ramos-Alvarado, Bladimir; Li Peiwen; Liu Hong; Hernandez-Guerrero, Abel

    2011-01-01

    A study of the heat transfer performance of liquid-cooled heat sinks with conventional and novel micro-channel flow field configurations for application in electronic devices, fuel cells, and concentrated solar cells is presented in this paper. The analyses were based on computations using the CFD software ANSYS FLUENT. The flow regime in heat sinks is constrained to laminar flow in the study. Details of the heat transfer performance, particularly, the uniformity of temperature distribution on the heating surface, as well as the pressure losses and pumping power in the operation of the studied heat sinks were obtained. Comparisons of the flow distribution uniformity in multiple flow channels, temperature uniformity on heating surfaces, and pumping power consumption of heat sinks with novel flow field configurations and conventional flow field configurations were conducted. It was concluded that the novel flow field configurations studied in this work exhibit appreciable benefits for application in heat sinks. - Highlights: → We present novel designs of flow channel configurations in liquid cooled heat sinks. → The flow and heat transfer in heat sinks were simulated using CFD tool. → The temperature and pressure loss in novel and conventional heat sinks were studied. → Figure of merit of heat sinks in different flow channel configurations was presented. → The heat sinks having our novel design of flow channel configurations are excellent.

  16. Test facility for auxiliary cooling system (ACS) of fast breeder reactor for Power Reactor and Nuclear Fuel Development Corporation (PNC)

    International Nuclear Information System (INIS)

    1983-01-01

    In preparation of constructing ''Monju'', a prototype fast breeder reactor, PNC has been pushing forward its research and development projects and the ACS was constructed under these projects. The auxiliary cooling system is an important engineered safety feature, and is used for safe removal of heat from the reactor at the shutdown. The ACS serves as a means of testing and assessing the auxiliary cooling system for the ''Monju'' and is designed and manufactured to have one fifth capacity of the Monju. The air heat exchanger and the ACS system was designed to withstand higher temperature range of the conventional design code (MITI-501), and finned tubes were applied for effective heat removal. Preheating system was designed to heat up the whole system over 200 0 C within 20 hours to prevent sodium from freezing. Basic performance of ACS was verified satisfactorily by a series of performance tests, such as start up test, flow rate measurement and preheating test before delivery. The experience from designing and construction of ACS and data obtained by these tests will be very instructive for designing and construction of the ''Monju''. (author)

  17. Impact of reducing sodium void worth on the severe accident response of metallic-fueled sodium-cooled reactors

    International Nuclear Information System (INIS)

    Wigeland, R.A.; Turski, R.B.; Pizzica, P.A.

    1994-01-01

    Analyses have performed on the severe accident response of four 90 MWth reactor cores, all designed using the metallic fuel of the Integrated Fast Reactor (IFR) concept. The four core designs have different sodium void worth, in the range of -3$ to 5$. The purpose of the investigation is to determine the improvement in safety, as measured by the severe accident consequences, that can be achieved from a reduction in the sodium void worth for reactor cores designed using the IFR concept

  18. Study on velocity field in a wire wrapped fuel pin bundle of sodium cooled reactor. Detailed velocity distribution in a subchannel

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Kobayashi, Jun; Miyakoshi, Hiroyuki; Kamide, Hideki

    2009-01-01

    A sodium cooled fast reactor is designed to attain a high burn-up core in a feasibility study on commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, deformation of fuel pin due to the swelling and thermal bowing may decrease local flow velocity via change of flow area in the subassembly and influence the heat removal capability. Therefore, it is of importance to obtain the flow velocity distribution in a wire wrapped pin bundle. A 2.5 times enlarged 7-pin bundle water model was applied to investigate the detailed velocity distribution in an inner subchannel surrounded by 3 pins with wrapping wire. The test section consisted of a hexagonal acrylic duct tube and fluorinated resin pins which had nearly the same refractive index with that of water and a high light transmission rate. The velocity distribution in an inner subchannel with the wrapping wire was measured by PIV (Particle Image Velocimetry) through the front and lateral sides of the duct tube. In the vertical velocity distribution in a narrow space between the pins, the wrapping wire decreased the velocity downstream of the wire and asymmetric flow distribution was formed between the pin and wire. In the horizontal velocity distribution, swirl flow around the wrapping wire was obviously observed. The measured velocity data are useful for code validation of pin bundle thermalhydraulics. (author)

  19. Updating of adventitious fuel pin failure frequency in sodium-cooled fast reactors and probabilistic risk assessment on consequent severe accident in Monju

    International Nuclear Information System (INIS)

    Fukano, Yoshitaka; Kurisaka, Kenichi; Nishimura, Masahiro; Naruto, Kenichi

    2015-01-01

    Experimental studies, deterministic approaches and probabilistic risk assessments (PRAs) on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious-fuel-pin-failures (AFPFs) have been considered to be the most dominant initiators of LFs in these PRAs because of their high frequency of occurrence during reactor operation and possibility of fuel-element-failure-propagation (FEFP). A PRA on FEFP from AFPF (FEFPA) in the Japanese prototype SFR (Monju) was performed in this study based on the state-of-the-art knowledge, reflecting the most recent operation procedures under off-normal conditions. Frequency of occurrence of AFPF in SFRs which was the initiating event of the event tree in this PRA was updated using a variety of methods based on the above-mentioned latest review on experiences of this phenomenon. As a result, the frequency of occurrence of, and the core damage frequency (CDF) from, AFPF in Monju was significantly reduced to a negligible magnitude compared with those in the existing PRAs. It was, therefore concluded that the CDF of FEFPA in Monju could be comprised in that of anticipated transient without scram or protected loss of heat sink events from both the viewpoint of occurrence probability and consequences. (author)

  20. Development of a brazing process for the production of water- cooled bipolar plates made of chromium-coated metal foils for PEM fuel cells

    International Nuclear Information System (INIS)

    Mueller, M; Hoehlich, D; Scharf, I; Lampke, T; Hollaender, U; Maier, H J

    2016-01-01

    Beside lithium batteries, PEM fuel cells are the most promising strategy as a power source to achieve the targets for introducing and increasing the usage of electric vehicles. Due to limited space and weight problems, water cooled, metallic bipolar plates in a fuel cell metal stack are preferred in motor vehicles. These plates are stamped metal sheets with a complex structure, interconnected media-tight. To meet the multiple tasks and requirements in use, complex and expensive combinations of materials are currently in use (carbon fiber composites, graphite, gold-plated nickel, stainless and acid resistant steel). The production of such plates is expensive as it is connected with considerable effort or the usage of precious metals. As an alternative, metalloid nitrides (CrN, VN, W 2 N, etc.) show a high chemical resistance, hardness and a good conductivity. So this material category meets the basic requirements of a top layer. However, the standard methods for their production (PVD, CVD) are expensive and have a slow deposition rate and a lower layer thicknesses. Because of these limitations, a full functionality over the life cycle of a bipolar plate is not guaranteed. The contribution shows the development and quantification of an alternative production process for bipolar plates. The expectation is to get significant advantages from the combination of chromium electrodeposition and thermochemical treatment to form chromium nitrides. Both processes are well researched and suitable for series production. The thermochemical treatment of the chromium layer also enables a process-integrated brazing. (paper)

  1. Spent nuclear fuel project cold vacuum drying facility tempered water and tempered water cooling system design description

    International Nuclear Information System (INIS)

    IRWIN, J.J.

    1998-01-01

    This document provides the System Design Description (SDD) for the Cold Vacuum Drying Facility (CVDF) Tempered Water (TW) and Tempered Water Cooling (TWC) System . The SDD was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998), The HNF-SD-SNF-DRD-O02, 1998, Cold Vacuum Drying Facility Design Requirements, and the CVDF Design Summary Report. The SDD contains general descriptions of the TW and TWC equipment, the system functions, requirements and interfaces. The SDD provides references for design and fabrication details, operation sequences and maintenance. This SOD has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved

  2. The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant

    Energy Technology Data Exchange (ETDEWEB)

    Ariani, Menik, E-mail: menikariani@gmail.com; Satya, Octavianus Cakra; Monado, Fiber [Department of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University, jl Palembang-Prabumulih km 32 Indralaya OganIlir, South of Sumatera (Indonesia); Su’ud, Zaki [Nuclear and Biophysics Research Division, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, jlGanesha 10, Bandung (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, 2-12-11N1-17 Ookayama, Meguro-Ku, Tokyo (Japan)

    2016-03-11

    The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactors with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on “Region-8” and “Region-10” core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years).

  3. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: fabrication of high-temperature gas-cooled reactor fuel containing uranium-233 and thorium

    International Nuclear Information System (INIS)

    Roddy, J.W.; Blanco, R.E.; Hill, G.S.; Moore, R.E.; Seagren, R.D.; Witherspoon, J.P.

    1976-06-01

    A cost/benefit study was made to determine the cost and effectiveness of various radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from model High-Temperature Gas-Cooled (HTGR) fuel fabrication plants and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term ''as low as reasonably achievable'' as it applies to these nuclear facilities. The base cases of the two model plants, a fresh fuel fabrication plant and a refabrication plant, are representative of current proposed commercial designs or are based on technology that is being developed to fabricate uranium, thorium, and graphite into fuel elements. The annual capacities of the fresh fuel plant and the refabrication plant are 450 and 245 metric tons of heavy metal (where heavy metal is uranium plus thorium), as charged to about fifty 1000-MW(e) HTGRs. Additional radwaste treatment systems are added to the base case plants in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The capital and annual costs for the added waste treatment operations and the corresponding reductions in dose commitments are calculated for each case. In the final analysis, the cost/benefit of each case, calculated as additional cost of radwaste system divided by the reduction in dose commitment, is tabulated or the dose commitment is plotted with cost as the variable. The status of each of the radwaste treatment methods is discussed. 48 figures, 74 tables

  4. Liquid metal cooled nuclear reactor

    International Nuclear Information System (INIS)

    Leigh, K.M.

    1980-01-01

    A liquid metal cooled nuclear reactor is described, wherein coolant is arranged to be flowed upwardly through a fuel assembly and having one or more baffles located above the coolant exit of the fuel assembly, the baffles being arranged so as to convert the upwardly directed motion of liquid metal coolant leaving the fuel assembly into a substantially horizontal motion. (author)

  5. Influence of fuel pin bowing on the temperature distribution in fuel pin cladding tubes in case of sodium cooling; experimental results

    International Nuclear Information System (INIS)

    Moeller, R.; Tschoeke, H.; Kolodziej, M.

    1978-09-01

    The influence of rod bowing on the local temperature distribution was measured with turbulent sodium flow in the cladding tubes of a 19-rod bundle mock-up of the SNR 300 Mark Ia fuel element. Such measurements have been carried out for the first time. The results presented in this report are part 1 of the experimental evaluation not yet completed. The major results are: 1. When a rod on the first ring gets deformed towards a neighbour on the second ring with a gap reduction from the nominal value of 100 % down to 20 %, the maximum azimuthal temperature difference of the outer rod increases by about 60 %. 2. The maximum azimuthal temperature difference of a rod on the first ring increases by a factor of 2, if it is approached by a neighbour on the same ring. 3. The reduction in cross section of a subchannel by rod bowing results only locally in distinct temperature rises, i.e. in the adjacent cladding tubes. Rods of the next but one row are no more subject to noticeable changes in temperature [de

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

    International Nuclear Information System (INIS)

    Govindha Rasu, N.; Velusamy, K.; Sundararajan, T.; Chellapandi, P.

    2013-01-01

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

  7. Control of microstructure in brazed zone of Zircaloy-4 nuclear fuel sheathing by optimization of Σ(C+P+Si) contents and cooling schedules

    International Nuclear Information System (INIS)

    Quach, V.; Northwood, D.O.

    1985-01-01

    In the production of fuel elements for the CANDU-PHW reactor, induction brazing is used to attach appendages (bearing and split spacer pads) onto the outside wall of the Zircaloy-4 sheathing. The brazing process, 40 to 60 seconds at temperature in excess of 1000 0 C, produces 3 heat-affected zones amounting to about 30% of the thickness. These heat affected zones quite often contain large grains and either a basketweave or a parallel plate type of Widmanstatten structure. Small grains and a basketweave structure are preferred. Using simulated brazing treatments, it is demonstrated that by control of the impurity content, Σ(C+P+Si), and cooling rate from the brazing temperature, the desired microstructure can be obtained in the braze heat-affected zone. The formation of the basketweave structure is promoted by higher impurity contents, with the second phase impurity particles acting as nuclei for the basketweave structure in preference to the β-grain boundaries where the parallel plate structure is nucleated

  8. Model-based control of a fuel cell cooling circuit with automatic software testing; Modellbasierte Steuerung des Kuehlkreislaufes einer Brennstoffzelle mit automatisiertem Test der Software

    Energy Technology Data Exchange (ETDEWEB)

    Schaefer, Sascha

    2012-07-01

    For reconstruction and control of the volume flow in a cooling circuit a fuel cell system is analyzed, and physical models of the fluid temperature and pump volume flow are derived. On this basis, functional models for derivation of software algorithms are presented which enable model-based calculation of the volume flow in a fluid circuit and coolant pump control on the basis of detailed system modelling. The available functions enable complete control and diagnosis of the volume flow. The functional models in Matlab/Simulink will be implemented in a control unit; they are discussed in the context of a software development process. It is stressed that correctness of the functions must be assured, and a trial automation system for functional models of this type is presented. (orig./AKB) [German] Zur Rekonstruktion und Steuerung des Volumenstroms in einem Kuehlkreislauf, wird ein Brennstoffzellensystem analysiert und physikalische Modelle fuer die Fluidtemperatur und den Pumpenvolumenstrom werden hergeleitet. Basierend auf diesen Zusammenhaengen werden Funktionsmodelle zur Ableitung von Softwarealgorithmen vorgestellt. Diese ermoeglichen es den Volumenstrom in einem Fluidkreislauf modellbasiert zu bestimmen bzw. die Kuehlmittelpumpe basierend auf einer detaillierten Systemmodellierung zu steuern. Die zur Verfuegung stehenden Funktionen erlauben eine komplette Regelung, Steuerung und Diagnose des Volumenstroms. Die Funktionsmodelle, in Matlab/Simulink realisiert und zur Ausfuehrung auf einem Steuergeraet vorgesehen, werden im Kontext eines Softwareentwicklungsprozesses diskutiert. Es wird die Notwendigkeit der Sicherstellung der Korrektheit der Funktionen herausgearbeitet und ein Testautomatisierungssystem fuer solche Funktionsmodelle vorgestellt.

  9. Development of new ORIGEN2 data library sets for research reactors with light water cooled oxide and silicide LEU (20 w/o) fuels based on JENDL-3.3 nuclear data

    International Nuclear Information System (INIS)

    Liem, Peng Hong; Sembiring, Tagor Malem

    2013-01-01

    Highlights: • We developed new ORIGEN2 data library sets for research reactors based on JENDL-3.3. • The sets cover oxide and silicide LEU fuels with meat density up to 4.74 g U/cm 3 . • Two kinds of data library sets are available: fuel region and non-fuel regions. • We verified the new data library sets with other codes. • We validated the new data library against a non-destructive test. -- Abstract: New sets of ORIGEN2 data library dedicated to research/testing reactors with light water cooled oxide and silicide LEU fuel plates based on JENDL-3.3 nuclear data were developed, verified and validated. The new sets are considered to be an extension of the most recent release of ORIGEN2.2UPJ code, i.e. the ORLIBJ33 library sets. The newly generated ORIGEN2 data library sets cover both oxide and silicide LEU fuels with fuel meat density range from 2.96 to 4.74 g U/cm 3 used in the present and future operation of the Indonesian 30 MWth RSG GAS research reactor. The new sets are expected applicable also for other research/testing reactors which utilize similar fuels or have similar neutron spectral indices. In addition to the traditional ORIGEN2 library sets for fuel depletion analyses in fuel regions, in the new data library sets, new ORIGEN2 library sets for irradiation/activation analyses were also prepared which cover all representative non-fuel regions of RSG GAS such as reflector elements, irradiation facilities, etc. whose neutron spectra are significantly softer than fuel regions. Verification with other codes as well as validation with a non-destructive test result showed promising results where a good agreement was confirmed

  10. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: reprocessing of high-temperature gas-cooled reactor fuel containing U-233 and thorium

    International Nuclear Information System (INIS)

    Davis, W. Jr.; Blanco, R.E.; Finney, B.C.; Hill, G.S.; Moore, R.E.; Witherspoon, J.P.

    1976-05-01

    A cost/benefit study was made to determine the cost and effectiveness of various radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from a model high-temperature gas-cooled reactor (HTGR) fuel reprocessing plant and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist the U. S. Nuclear Regulatory Commission in defining the term as low as reasonably achievable as it applies to this nuclear facility. The base case is representative of conceptual, developing technology of head-end graphite-burning operations and of extensions of solvent-extraction technology of current designs for light-water-reactor (LWR) fuel reprocessing plants. The model plant has an annual capacity of 450 metric tons of heavy metal (MTHM, where heavy metal is uranium plus thorium), as charged to about fifty 1000-MW(e) HTGRs. Additional radwaste treatment systems are added to the base-case plant in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The capital and annual costs for the added waste treatment operations and the corresponding reductions in dose commitments are calculated for each case. In the final analysis, the cost/benefit of each case, calculated as additional cost of radwaste system divided by the reduction in dose commitment, is tabulated or the dose commitment is plotted with cost as the variable. The status of each of the radwaste treatment methods used in the case studies is discussed

  11. Cooling towers

    International Nuclear Information System (INIS)

    Boernke, F.

    1975-01-01

    The need for the use of cooling systems in power plant engineering is dealt with from the point of view of a non-polluting form of energy production. The various cooling system concepts up to the modern natural-draught cooling towers are illustrated by examples. (TK/AK) [de

  12. Analysis of the thorium inclusion in the fuel of a fast reactor cooled by lead; Analisis de la inclusion de torio en el combustible de un reactor rapido enfriado por plomo

    Energy Technology Data Exchange (ETDEWEB)

    Juarez M, L. C.; Francois L, J. L., E-mail: lcjm@ier.unam.mx [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico)

    2017-09-15

    In the present work, we first verified a model of the European reactor cooled with lead (ELFR). The calculations were made with the code Monte Carlo serpent 2.27 and the library of cross sections Jeff-3.1. For this verification, three neutron parameters were compared: the evolution of the neutron multiplication factor, the Doppler constant and the effect of the vacuum fraction of the refrigerant, obtaining a good approximation with the reference values. Subsequently, the inclusion of thorium as a fertile material within the fuel was analyzed and the same neutron parameters were compared with the original fuel. The evolution of criticality for the case of thorium fuel differs significantly with respect to that of the original fuel (without thorium); this is due mainly to the breeding of the fissile isotope {sup 233}U. Therefore, is possible to have a longer fuel cycle, favoring the availability factor of the plant, without compromising the performance of the reactor since both the Doppler constant and the effect of the vacuum fraction of the refrigerant show a similar tendency to those of the original fuel, being negative in both cases. (Author)

  13. Influence of carbon monoxide on the cathode in high-temperature polymer electrolyte membrane fuel cells

    DEFF Research Database (Denmark)

    Søndergaard, Stine; Cleemann, Lars Nilausen; Jensen, Jens Oluf

    2017-01-01

    This paper describes the results of adding small amounts of CO gas to the cathode side in a HT-PEM fuel cell with a polybenzimidazole (PBI) membrane running on either oxygen or air. Experimental conditions: Temperature ranges 120–160 °C, constant current either 200 mA/cm2 or 800 mA/cm2 and CO...... improvement of the potential is seen before the situation goes back to normal. A good explanation for this is a competition between CO, O2 and H3PO4 at the three phase boundaries, also that a steady state exist in which CO constantly is oxidized to CO2....

  14. Sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hokkyo, N; Inoue, K; Maeda, H

    1968-11-21

    In a sodium cooled fast neutron reactor, an ultrasonic generator is installed at a fuel assembly hold-down mechanism positioned above a blanket or fission gas reservoir located above the core. During operation of the reactor an ultrsonic wave of frequency 10/sup 3/ - 10/sup 4/ Hz is constantly transmitted to the core to resonantly inject the primary bubble with ultrasonic energy to thereby facilitate its growth. Hence, small bubbles grow gradually to prevent the sudden boiling of sodium if an accident occurs in the cooling system during operation of the reactor.

  15. Nuclear fuels

    International Nuclear Information System (INIS)

    2008-01-01

    The nuclear fuel is one of the key component of a nuclear reactor. Inside it, the fission reactions of heavy atoms, uranium and plutonium, take place. It is located in the core of the reactor, but also in the core of the whole nuclear system. Its design and properties influence the behaviour, the efficiency and the safety of the reactor. Even if it represents a weak share of the generated electricity cost, its proper use represents an important economic stake. Important improvements remain to be made to increase its residence time inside the reactor, to supply more energy, and to improve its robustness. Beyond the economical and safety considerations, strategical questions have to find an answer, like the use of plutonium, the management of resources and the management of nuclear wastes and real technological challenges have to be taken up. This monograph summarizes the existing knowledge about the nuclear fuel, its behaviour inside the reactor, its limits of use, and its R and D tracks. It illustrates also the researches in progress and presents some key results obtained recently. Content: 1 - Introduction; 2 - The fuel of water-cooled reactors: aspect, fabrication, behaviour of UO 2 and MOX fuels inside the reactor, behaviour in loss of tightness situation, microscopic morphology of fuel ceramics and evolution under irradiation - migration and localisation of fission products in UOX and MOX matrices, modeling of fuels behaviour - modeling of defects and fission products in the UO 2 ceramics by ab initio calculations, cladding and assembly materials, pellet-cladding interaction, advanced UO 2 and MOX ceramics, mechanical behaviour of the fuel assembly, fuel during a loss of coolant accident, fuel during a reactivity accident, fuel during a serious accident, fuel management inside reactor cores, fuel cycle materials balance, long-term behaviour of the spent fuel, fuel of boiling water reactors; 3 - the fuel of liquid metal fast reactors: fast neutrons radiation

  16. Spray cooling

    International Nuclear Information System (INIS)

    Rollin, Philippe.

    1975-01-01

    Spray cooling - using water spraying in air - is surveyed as a possible system for make-up (peak clipping in open circuit) or major cooling (in closed circuit) of the cooling water of the condensers in thermal power plants. Indications are given on the experiments made in France and the systems recently developed in USA, questions relating to performance, cost and environmental effects of spray devices are then dealt with [fr

  17. Emergency core cooling systems

    International Nuclear Information System (INIS)

    Kubokoya, Takashi; Okataku, Yasukuni.

    1984-01-01

    Purpose: To maintain the fuel soundness upon loss of primary coolant accidents in a pressure tube type nuclear reactor by injecting cooling heavy water at an early stage, to suppress the temperature of fuel cans at a lower level. Constitution: When a thermometer detects the temperature rise and a pressure gauge detects that the pressure for the primary coolants is reduced slightly from that in the normal operation upon loss of coolant accidents in the vicinity of the primary coolant circuit, heavy water is caused to flow in the heavy water feed pipeway by a controller. This enables to inject the heavy water into the reactor core in a short time upon loss of the primary coolant accidents to suppress the temperature rise in the fuel can thereby maintain the fuel soundness. (Moriyama, K.)

  18. High performance fuel technology development

    Energy Technology Data Exchange (ETDEWEB)

    Koon, Yang Hyun; Kim, Keon Sik; Park, Jeong Yong; Yang, Yong Sik; In, Wang Kee; Kim, Hyung Kyu [KAERI, Daejeon (Korea, Republic of)

    2012-01-15

    {omicron} Development of High Plasticity and Annular Pellet - Development of strong candidates of ultra high burn-up fuel pellets for a PCI remedy - Development of fabrication technology of annular fuel pellet {omicron} Development of High Performance Cladding Materials - Irradiation test of HANA claddings in Halden research reactor and the evaluation of the in-pile performance - Development of the final candidates for the next generation cladding materials. - Development of the manufacturing technology for the dual-cooled fuel cladding tubes. {omicron} Irradiated Fuel Performance Evaluation Technology Development - Development of performance analysis code system for the dual-cooled fuel - Development of fuel performance-proving technology {omicron} Feasibility Studies on Dual-Cooled Annular Fuel Core - Analysis on the property of a reactor core with dual-cooled fuel - Feasibility evaluation on the dual-cooled fuel core {omicron} Development of Design Technology for Dual-Cooled Fuel Structure - Definition of technical issues and invention of concept for dual-cooled fuel structure - Basic design and development of main structure components for dual- cooled fuel - Basic design of a dual-cooled fuel rod.

  19. Water cooling system for sintering furnaces of nuclear fuel pellets; Sistema de enfriamiento con agua para hornos de sinterizado de pastillas de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-11-01

    This work has as a main objective to develop a continuous cooling water system, which is necessary for the cooling of the sintering furnaces. This system is used to protect them as well as for reducing the water consumption, ejecting the heat generated into this furnaces and scattering it into the atmosphere in a fast and continuous way. The problem was defined and the reference parameters established, making the adequate research. The materials were selected as well as the length of the pipeline which will carry the secondary refrigerant fluid (water). Three possible solutions were tried,and evaluated, and from these, the thermal and economically most efficient option was selected. The layout of the solution was established and the theoretical construction of a cooling system for liquids using dichlorofluoromethane (R-22), as a refrigerant and a air cooled condenser, was accomplished. (Author).

  20. Gas cooled high temperature reactor with a heap of pebble shaped fuel elements and absorber rods which can be driven directly into the heap of pebbles

    International Nuclear Information System (INIS)

    Elter, C.; Schmitt, H.; Schoening, J.; Weicht, U.

    1980-01-01

    The absorber rod for the graphite moderated, helium cooled reactor is cylindrical and has a tip in the shape of the frustrum of a cone. It consists of three coaxially arranged sleeve tubes made of steel, the inner and centre sleeve tubes surrounding the absorber part (B4C) so as to be gastight. The inner sleeve tube represents the supporting tube and is cooled by cold gas, as is the annular gap between the centre and outer sleeve tube. (RW) [de

  1. Nuclear fuel replacement device

    International Nuclear Information System (INIS)

    Ritz, W.C.; Robey, R.M.; Wett, J.F.

    1984-01-01

    A fuel handling arrangement for a liquid metal cooled nuclear reactor having a single rotating plug eccentric to the fuel core and a fuel handling machine radially movable along a slot in the plug with a transfer station disposed outside the fuel core but covered by the eccentric plug and within range of movement of said fuel handling machine to permit transfer of fuel assemblies between the core and the transfer station. (author)

  2. A study of the friction and wear processes of the structural components of fuel assemblies for water-cooled and water moderated power reactors

    International Nuclear Information System (INIS)

    Makarov, V.; Afanasiev, A.; Matvienko, I.; Drozdov, Y.; Puchkov, V.

    2011-01-01

    The friction forces affect the fuel assembly (FA) strength at all the stages of its lifecycle. The paper covers the methods and the results of the pre-irradiation experimental studies of the static and dynamic processes the friction forces are involved in. These comprise the FA assembling at the manufacturer, fuel rod flow-induced vibration and fretting-wear in the fuel rod-to-cell friction pairs, rod cluster control assembly (RCCA) movement in the FA guide tubes, FA bowing, FA loading-unloading into the core, irradiation-induced growth and thermal-mechanical fuel rod-to-spacer grid interaction. (authors)

  3. Evolution of the thickness of the aluminum oxide film due to the pH of the cooling water and surface temperature of the fuel elements clad of a nuclear reactor

    International Nuclear Information System (INIS)

    Babiche, Ivan

    2013-01-01

    This paper describes the mechanism of growth of a film of aluminum oxide on an alloy of the same material, which serves as a protective surface being the constituent material of the RP-10 nuclear reactor fuel elements clads. The most influential parameters on the growth of this film are: the pH of the cooling water and the clad surface temperature of the fuel element. For this study, a mathematical model relating the evolution of the aluminum oxide layer thickness over the time, according to the same oxide film using a power law is used. It is concluded that the time of irradiation, the heat flux at the surface of the aluminum material, the speed of the coolant, the thermal conductivity of the oxide, the initial thickness of the oxide layer and the solubility of the protective oxide are parameters affecting in the rate and film formation. (author).

  4. Aerodynamic and thermal studies of cans of gas cooled fuel elements; Etudes aerodynamiques et thermiques de gaines d'elements combustibles refroidis par gaz

    Energy Technology Data Exchange (ETDEWEB)

    Gelin, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Milliat, J P [Electricite de France (EDF), 75 - Paris (France)

    1964-07-01

    Research on clusters was undertaken at the CEA in 1959, while studies on herring-bone cans were developed at the EDF Laboratory at Chatou and the CEA laboratory at Saclay at the end of 1959. In 1962, a general study on corrugations was begun at the Saclay Laboratory with a view to improving the clusters, and continued later in both laboratories relative to the internal cooling of annular fuel elements. As these studies progressed, trial facilities were extended while experimental methods have improved constantly. At the present time, both laboratories, working in complete collaboration, have powerful means at their disposal. Work on the clusters has been concerned chiefly with pressure losses due to the assembly parts, and with the temperature variations around the elements of the cluster. In this way, we have been able to determine satisfactorily the hot points of the can, the deformations of the rods and the conditions of stability of these deformations. In the case of the herring-bone cans, studies have been directed to the evolution of performances as a function of the geometric parameters on the one hand, and to the special aerodynamic and thermal features caused by the fins and by interruptions of cartridge on the other hand. These studies have led to a very thorough knowledge of the cartridges chosen for the reactors EDF 2 and EDF 3, and now open up very hopeful prospects for future reactors, particularly those fitted with annular elements; among the alternatives suitable for the inner surface of the annular element can, corrugations and longitudinal fins have been fairly extensively tested over a wide range of Reynolds number. (authors) [French] Les etudes de grappes ont ete entreprises au C.E.A. en 1959, tandis que les etudes de gaines a chevrons se sont developpees au Laboratoire de Chatou de l'EDF et au Laboratoire de Saclay du C.E.A. a la fin de l'annee 1959. En 1962, une etude generale sur les corrugations a ete commencee au Laboratoire de Saclay pour

  5. Gas-cooled reactors

    International Nuclear Information System (INIS)

    Schulten, R.; Trauger, D.B.

    1976-01-01

    Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing

  6. Passive Safety Systems in Advanced Water Cooled Reactors (AWCRS). Case Studies. A Report of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    2013-09-01

    This report presents the results from the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) collaborative project (CP) on Advanced Water Cooled Reactor Case Studies in Support of Passive Safety Systems (AWCR), undertaken under the INPRO Programme Area C. INPRO was launched in 2000 - on the basis of a resolution of the IAEA General Conference (GC(44)/RES/21) - to ensure that nuclear energy is available in the 21st century in a sustainable manner, and it seeks to bring together all interested Member States to consider actions to achieve innovation. An important objective of nuclear energy system assessments is to identify 'gaps' in the various technologies and corresponding research and development (R and D) needs. This programme area fosters collaboration among INPRO Member States on selected innovative nuclear technologies to bridge technology gaps. Public concern about nuclear reactor safety has increased after the Fukushima Daiichi nuclear power plant accident caused by the loss of power to pump water for removing residual heat in the core. As a consequence, there has been an increasing interest in designing safety systems for new and advanced reactors that are passive in nature. Compared to active systems, passive safety features do not require operator intervention, active controls, or an external energy source. Passive systems rely only on physical phenomena such as natural circulation, thermal convection, gravity and self-pressurization. Passive safety features, therefore, are increasingly recognized as an essential component of the next-generation advanced reactors. A high level of safety and improved competitiveness are common goals for designing advanced nuclear power plants. Many of these systems incorporate several passive design concepts aimed at improving safety and reliability. The advantages of passive safety systems include simplicity, and avoidance of human intervention, external power or signals. For these reasons, most

  7. Nuclear fuels

    International Nuclear Information System (INIS)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

    2009-01-01

    fuel, Anticipated evolution of fuel in dry storage, Anticipated evolution of fuel in deep geological disposal); Boiling-water reactor fuel (Similarities, and differences with PWR fuel, Axial and radial zoning, Rod and channel box sizes, Poisoning and reactivity control, Cladding specific characteristics, Trends in fuel evolution); 3 - Liquid-metal-cooled fast reactor fuel: Fast-neutron irradiation damage in structural materials (Fast-neutron-induced damage in metals, What materials should be used?); Fuels and targets for fast-reactor transmutation (Fast reactors: reactors affording the ability to carry out effective actinide transmutation, Recycling: homogeneous, or heterogeneous?); 4 - gas-cooled reactor fuel: Particle fuel (From the initial concept to the advanced TRISO particle concept, Kernel fabrication processes, Particle coating by chemical vapor deposition, Fuel element fabrication: particle compaction, Characterization of fuel particles, and elements, From HTR fuel to VHTR and GFR fuels: the GAIA facility at CEA/Cadarache); Irradiation behavior of particle fuels (Particle fuel: a variety of failure modes for a high-strength object, The amoeba effect, Fission product behavior, and diffusion in particle fuels); Mechanical modeling of particle fuel; Very-high-temperature reactor (VHTR) fuel; Gas-cooled fast reactor (GFR) fuel (The specifications for GFR fuel, GFR fissile material, First containment baffler materials, GFR fuel element concepts); 5 - Research reactor fuels (A considerable feedback from experience, Conversion of French reactors to low-enriched (≤20% U-235)U 3 Si 2 fuel, Conversion of all reactors: R and D requirements for high-performance reactors, An 'advanced' research reactor fuel: UMo, The startup fuel for the Jules Horowitz Reactor (JHR) will still be U 3 Si 2 -Al; 6 - An instrument for future fuel research: the Jules Horowitz Reactor (JHR): Fuel irradiation experiments in JHR, JHR: a flexible instrument; 7 - Glossary-Index

  8. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    nature of spent nuclear fuel, Anticipated evolution of fuel in dry storage, Anticipated evolution of fuel in deep geological disposal); Boiling-water reactor fuel (Similarities, and differences with PWR fuel, Axial and radial zoning, Rod and channel box sizes, Poisoning and reactivity control, Cladding specific characteristics, Trends in fuel evolution); 3 - Liquid-metal-cooled fast reactor fuel: Fast-neutron irradiation damage in structural materials (Fast-neutron-induced damage in metals, What materials should be used?); Fuels and targets for fast-reactor transmutation (Fast reactors: reactors affording the ability to carry out effective actinide transmutation, Recycling: homogeneous, or heterogeneous?); 4 - gas-cooled reactor fuel: Particle fuel (From the initial concept to the advanced TRISO particle concept, Kernel fabrication processes, Particle coating by chemical vapor deposition, Fuel element fabrication: particle compaction, Characterization of fuel particles, and elements, From HTR fuel to VHTR and GFR fuels: the GAIA facility at CEA/Cadarache); Irradiation behavior of particle fuels (Particle fuel: a variety of failure modes for a high-strength object, The amoeba effect, Fission product behavior, and diffusion in particle fuels); Mechanical modeling of particle fuel; Very-high-temperature reactor (VHTR) fuel; Gas-cooled fast reactor (GFR) fuel (The specifications for GFR fuel, GFR fissile material, First containment baffler materials, GFR fuel element concepts); 5 - Research reactor fuels (A considerable feedback from experience, Conversion of French reactors to low-enriched ({<=}20% U-235)U{sub 3}Si{sub 2} fuel, Conversion of all reactors: R and D requirements for high-performance reactors, An 'advanced' research reactor fuel: UMo, The startup fuel for the Jules Horowitz Reactor (JHR) will still be U{sub 3}Si{sub 2}-Al; 6 - An instrument for future fuel research: the Jules Horowitz Reactor (JHR): Fuel irradiation experiments in JHR, JHR: a flexible

  9. Ventilative Cooling

    DEFF Research Database (Denmark)

    Heiselberg, Per Kvols; Kolokotroni, Maria

    This report, by venticool, summarises the outcome of the work of the initial working phase of IEA ECB Annex 62 Ventilative Cooling and is based on the findings in the participating countries. It presents a summary of the first official Annex 62 report that describes the state-of-the-art of ventil......This report, by venticool, summarises the outcome of the work of the initial working phase of IEA ECB Annex 62 Ventilative Cooling and is based on the findings in the participating countries. It presents a summary of the first official Annex 62 report that describes the state......-of-the-art of ventilative cooling potentials and limitations, its consideration in current energy performance regulations, available building components and control strategies and analysis methods and tools. In addition, the report provides twenty six examples of operational buildings using ventilative cooling ranging from...

  10. Fuel assembly

    International Nuclear Information System (INIS)

    Nomata, Terumitsu.

    1993-01-01

    Among fuel pellets to be loaded to fuel