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Sample records for graphite fuel pebbles

  1. 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.

  2. Abrasion behavior of graphite pebble in lifting pipe of pebble-bed HTR

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Ke; Su, Jiageng [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Zhou, Hongbo [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Chinergy Co., LTD., Beijing 100193 (China); Peng, Wei; Liu, Bing [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Yu, Suyun, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Educations, Tsinghua University, Beijing 10084 (China)

    2015-11-15

    Highlights: • Quantitative determination of abrasion rate of graphite pebbles in different lifting velocities. • Abrasion behavior of graphite pebble in helium, air and nitrogen. • In helium, intensive collisions caused by oscillatory motion result in more graphite dust production. - Abstract: A pebble-bed high-temperature gas-cooled reactor (pebble-bed HTR) uses a helium coolant, graphite core structure, and spherical fuel elements. The pebble-bed design enables on-line refueling, avoiding refueling shutdowns. During circulation process, the pebbles are lifted pneumatically via a stainless steel lifting pipe and reinserted into the reactor. Inevitably, the movement of the fuel elements as they recirculate in the reactor produces graphite dust. Mechanical wear is the primary source of graphite dust production. Specifically, the sources are mechanisms of pebble–pebble contact, pebble–wall (structural graphite) contact, and fuel handling (pebble–metal abrasion). The key contribution to graphite dust production is from the fuel handling system, particularly from the lifting pipe. During pneumatic lift, graphite pebbles undergo multiple collisions with the stainless steel lifting pipe, thereby causing abrasion of the graphite pebbles and producing graphite dust. The present work explored the abrasion behavior of graphite pebble in the lifting pipe by measuring the abrasion rate at different lifting velocities. The abrasion rate of the graphite pebble in helium was found much higher than those in air and nitrogen. This gas environment effect could be explained by either tribology behavior or dynamic behavior. Friction testing excluded the possibility of tribology reason. The dynamic behavior of the graphite pebble was captured by analysis of the audio waveforms during pneumatic lift. The analysis results revealed unique dynamic behavior of the graphite pebble in helium. Oscillation and consequently intensive collisions occur during pneumatic lift, causing

  3. Effect of a Central Graphite Column on a Pebble Flow in a Pebble Bed Core

    International Nuclear Information System (INIS)

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

    2006-01-01

    A pebble bed reactor(PBR) uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. The pebble bed core is configured as cylindrical or annular depending on the reactor power. It is well known that an annular core can increase a cores' thermal power. The annular inner core zone is typically filled with movable graphite balls or a fixed graphite column. The first problem with this conventional annular core is that it is difficult to maintain a boundary between the central graphite ball zone and the outer fuel zone. The second problem is that it is expensive to replace the central fixed graphite column after several tens of years of reactor operation. In order to resolve these problems, a PBR with a central graphite column in a low core is invented. This paper presents the effect of the central graphite column on a pebble flow by using the computational fluid dynamics(CFD) code, CFX-10

  4. Surface coating of graphite pebbles for Korean HCCR TBM

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youngmin [National Fusion Research Institute, Daejeon (Korea, Republic of); Yun, Young-Hoon, E-mail: yunh2@dsu.ac.kr [Dongshin University, Naju (Korea, Republic of); Park, Yi-Hyun; Ahn, Mu-Young; Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Highlights: • A CVR-SiC coating was successfully formed on graphite pebbles for neutron reflector. • Dense and fine-grained surface morphologies of the SiC coatings were observed. • Oxidation resistance of the CVR-SiC-coated graphite pebbles was improved. - Abstract: The new concept of the recently modified Helium-Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) is to adopt a graphite reflector in the form of a pebble bed. A protective SiC coating is applied to the graphite pebbles to prohibit their reaction with steam or air as well as dust generation during TBM operation. In this research, the chemical vapor reaction (CVR) method was applied to fabricate SiC-coated graphite pebbles in a silica source. Relatively dense CVR-SiC coating was successfully formed on the graphite pebbles through the reduction of the graphite phase with SiO gas that was simply created from the silica source at 1850 °C (2 h). The microstructural features, XRD patterns, pore-size distribution and oxidation behavior of the SiC-coated graphite pebbles were investigated. To develop the practical process, which will be applied for mass production hereafter, a novel alternative method was applied to form the layer of SiC coating on the graphite pebbles over the silica source.

  5. Surface coating of graphite pebbles for Korean HCCR TBM

    International Nuclear Information System (INIS)

    Lee, Youngmin; Yun, Young-Hoon; Park, Yi-Hyun; Ahn, Mu-Young; Cho, Seungyon

    2014-01-01

    Highlights: • A CVR-SiC coating was successfully formed on graphite pebbles for neutron reflector. • Dense and fine-grained surface morphologies of the SiC coatings were observed. • Oxidation resistance of the CVR-SiC-coated graphite pebbles was improved. - Abstract: The new concept of the recently modified Helium-Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) is to adopt a graphite reflector in the form of a pebble bed. A protective SiC coating is applied to the graphite pebbles to prohibit their reaction with steam or air as well as dust generation during TBM operation. In this research, the chemical vapor reaction (CVR) method was applied to fabricate SiC-coated graphite pebbles in a silica source. Relatively dense CVR-SiC coating was successfully formed on the graphite pebbles through the reduction of the graphite phase with SiO gas that was simply created from the silica source at 1850 °C (2 h). The microstructural features, XRD patterns, pore-size distribution and oxidation behavior of the SiC-coated graphite pebbles were investigated. To develop the practical process, which will be applied for mass production hereafter, a novel alternative method was applied to form the layer of SiC coating on the graphite pebbles over the silica source

  6. PBMR Project - Pebble Fuel Advantages

    International Nuclear Information System (INIS)

    Slabber, Johan; Matzie, Regis; Casperson, Sten; Kriel, Willem

    2006-01-01

    An overview is presented of all the important issues that influenced the choice of pebble fuel for the High-temperature Gas-cooled Reactor (HTGR) concept developed by South Africa. Each of these issues is then discussed in detail and compared with other fuel configurations proposed for direct cycle High-temperature Reactor (HTR) applications. The comparisons are provided using objective data generated by analyses done for the design of the Pebble Bed Modular Reactor (PBMR) and data that is available in open literature for the other fuel configurations

  7. Transient heat conduction in a pebble fuel applying fractional model

    International Nuclear Information System (INIS)

    Gomez A, R.; Espinosa P, G.

    2009-10-01

    In this paper we presents the equation of thermal diffusion of temporary-fractional order in the one-dimensional space in spherical coordinates, with the objective to analyze the heat transference between the fuel and coolant in a fuel element of a Pebble Bed Modular Reactor. The pebble fuel is the heterogeneous system made by microsphere constitutes by U O, pyrolytic carbon and silicon carbide mixed with graphite. To describe the heat transfer phenomena in the pebble fuel we applied a constitutive law fractional (Non-Fourier) in order to analyze the behaviour transient of the temperature distribution in the pebble fuel with anomalous thermal diffusion effects a numerical model is developed. (Author)

  8. Loads on pebble bed fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Teuchert, E.; Maly, V.

    1974-03-15

    A comparison is made of key parameters for multi-recycle pebbles and single-pass once-through (OTTO) pebbles. The parameters analyzed include heat transfer characteristics with burn-up, temperature profiles, power per element as a function of axial position in the core, and burn-up. For the OTTO-scheme, the comparisons addressed the use of the conventional fuel element and the advanced "shell ball" designed to reduce the peak fuel temperature in the center of the fuel element. All studies addressed the uranium-thorium fuel cycle.

  9. Verification of two-temperature method for heat transfer process within a pebble fuel

    International Nuclear Information System (INIS)

    Yu Dali; Peng Minjun

    2014-01-01

    A typical pebble fuel that used in high temperature reactor (HTR), mainly consists of a graphite matrix with numerous dispersed tristructural-isotropic (TRISO) fuel particles and a surrounding thin non-fueled graphite shell. These high heterogeneities lead to difficulty in explicit thermal calculation of a pebble fuel. We proposed a two-temperature method (TTM) to calculate the temperature distribution within a pebble fuel. The method is not only convenient to perform but also gives more realistic results since particles and graphite matrix are considered separately while the traditional ways are considering the fuel zone as average heat generation source. The method is validated both by Computational Fluid Dynamics (CFD) method and Wiener bounds. Results show that TTM has a stable performance and high accuracy. (author)

  10. Pebble Bed Reactor: core physics and fuel cycle analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Worley, B.A.

    1979-10-01

    The Pebble Bed Reactor is a gas-cooled, graphite-moderated high-temperature reactor that is continuously fueled with small spherical fuel elements. The projected performance was studied over a broad range of reactor applicability. Calculations were done for a burner on a throwaway cycle, a converter with recycle, a prebreeder and breeder. The thorium fuel cycle was considered using low, medium (denatured), and highly enriched uranium. The base calculations were carried out for electrical energy generation in a 1200 MW/sub e/ plant. A steady-state, continuous-fueling model was developed and one- and two-dimensional calculations were used to characterize performance. Treating a single point in time effects considerable savings in computer time as opposed to following a long reactor history, permitting evaluation of reactor performance over a broad range of design parameters and operating modes.

  11. Development and testing of nuclear graphite for the German pebble-bed high temperature reactor

    International Nuclear Information System (INIS)

    Haag, G.; Delle, W.; Nickel, H.; Theymann, W.; Wilhelmi, G.

    1987-01-01

    Several types of high temperature reactors have been developed in the Federal Republic of Germany. They are all based on spherical fuel elements being surrounded by graphite as reflector material. As an example, HTR-500 developed by the Hochtemperatur Reaktorbau GmbH is shown. The core consists of the top reflector, the side reflector with inner and outer parts, the bottom reflector and the core support columns. The most serious problem with respect to fast neutron radiation damage had to be solved for the materials of those parts near the pebble bed. Regarding the temperature profile in the core, the top reflector is at 300 deg C, and as cooling gas flows from the top downward, the temperature of the inner side reflector rises to about 700 deg C at the bottom. Fortunately, the highest fast neutron load accumulated during the life time of a reactor corresponds to the lowest temperature. This makes graphite components easier to survive neutron exposure without being mechanically damaged, although the maximum fast neutron fluence is as high as 4 x 10 22 /cm 2 at about 400 deg C. HTR graphite components are divided into four classes according to loading. The raw materials for nuclear graphite, the development of pitch coke nuclear graphite, the irradiation behavior of ATR-2E and ASR-IRS and others are reported. (Kako, I.)

  12. Optimization of MOX fuel cycles in pebble bed HTGR

    International Nuclear Information System (INIS)

    Wei Jinfeng; Li Fu; Sun Yuliang

    2013-01-01

    Compared with light water reactor (LWR), the pebble bed high temperature gas-cooled reactor (HTGR) is able to operate in a full mixed oxide (MOX) fuelled core without significant change to core structure design. Based on a reference design of 250 MW pebble bed HTGR, four MOX fuel cycles were designed and evaluated by VSOP program package, including the mixed Pu-U fuel pebbles and mixed loading of separate Pu-pebbles and U-pebbles. Some important physics features were investigated and compared for these four cycles, such as the effective multiplication factor of initial core, the pebble residence time, discharge burnup, and temperature coefficients. Preliminary results show that the overall performance of one case is superior to other equivalent MOX fuel cycles on condition that uranium fuel elements and plutonium fuel elements are separated as the different fuel pebbles and that the uranium fuel elements are irradiated longer in the core than the plutonium fuel elements, and the average discharge burnup of this case is also higher than others. (authors)

  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. HTR-proteus pebble bed experimental program core 4: random packing with a 1:1 moderator-to-fuel pebble ratio

    Energy Technology Data Exchange (ETDEWEB)

    Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sterbentz, James W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Briggs, J. Blair [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Snoj, Luka [Jozef Stefan Inst. (IJS), Ljubljana (Slovenia); Lengar, Igor [Jozef Stefan Inst. (IJS), Ljubljana (Slovenia); Koberl, Oliver [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    2014-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering

  16. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORE 4: RANDOM PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Leland M. Montierth

    2013-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering

  17. HTR-PROTEUS Pebble Bed Experimental Program Cores 1, 1A, 2, and 3: Hexagonal Close Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Barbara H. Dolphin; James W. Sterbentz; Luka Snoj; Igor Lengar; Oliver Köberl

    2013-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. Four benchmark experiments were evaluated in this report: Cores 1, 1A, 2, and 3. These core configurations represent the hexagonal close packing (HCP) configurations of the HTR-PROTEUS experiment with a moderator-to-fuel pebble ratio of 1:2. Core 1 represents the only configuration utilizing ZEBRA control rods. Cores 1A, 2, and 3 use withdrawable, hollow, stainless steel control rods. Cores 1 and 1A are similar except for the use of different control rods; Core 1A also has one less layer of pebbles (21 layers instead of 22). Core 2 retains the first 16 layers of pebbles from Cores 1 and 1A and has 16 layers of moderator pebbles stacked above the fueled layers. Core 3 retains the first 17 layers of pebbles but has polyethylene rods inserted between pebbles to simulate water ingress. The additional partial pebble layer (layer 18) for Core 3 was not included as it was used for core operations and not the reported critical configuration. Cores 1, 1A, 2, and 3 were determined to be acceptable benchmark experiments.

  18. HTR-PROTEUS Pebble Bed Experimental Program Cores 1, 1A, 2, and 3: Hexagonal Close Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Barbara H. Dolphin; James W. Sterbentz; Luka Snoj; Igor Lengar; Oliver Köberl

    2012-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. Four benchmark experiments were evaluated in this report: Cores 1, 1A, 2, and 3. These core configurations represent the hexagonal close packing (HCP) configurations of the HTR-PROTEUS experiment with a moderator-to-fuel pebble ratio of 1:2. Core 1 represents the only configuration utilizing ZEBRA control rods. Cores 1A, 2, and 3 use withdrawable, hollow, stainless steel control rods. Cores 1 and 1A are similar except for the use of different control rods; Core 1A also has one less layer of pebbles (21 layers instead of 22). Core 2 retains the first 16 layers of pebbles from Cores 1 and 1A and has 16 layers of moderator pebbles stacked above the fueled layers. Core 3 retains the first 17 layers of pebbles but has polyethylene rods inserted between pebbles to simulate water ingress. The additional partial pebble layer (layer 18) for Core 3 was not included as it was used for core operations and not the reported critical configuration. Cores 1, 1A, 2, and 3 were determined to be acceptable benchmark experiments.

  19. Pebble bed pebble motion: Simulation and Application

    Science.gov (United States)

    Cogliati, Joshua J.

    Pebble bed reactors (PBR) have moving graphite fuel pebbles. This unique feature provides advantages, but also means that simulation of the reactor requires understanding the typical motion and location of the granular flow of pebbles. This dissertation presents a method for simulation of motion of the pebbles in a PBR. A new mechanical motion simulator, PEBBLES, efficiently simulates the key elements of motion of the pebbles in a PBR. This model simulates gravitational force and contact forces including kinetic and true static friction. It's used for a variety of tasks including simulation of the effect of earthquakes on a PBR, calculation of packing fractions, Dancoff factors, pebble wear and the pebble force on the walls. The simulator includes a new differential static friction model for the varied geometries of PBRs. A new static friction benchmark was devised via analytically solving the mechanics equations to determine the minimum pebble-to-pebble friction and pebble-to-surface friction for a five pebble pyramid. This pyramid check as well as a comparison to the Janssen formula was used to test the new static friction equations. Because larger pebble bed simulations involve hundreds of thousands of pebbles and long periods of time, the PEBBLES code has been parallelized. PEBBLES runs on shared memory architectures and distributed memory architectures. For the shared memory architecture, the code uses a new O(n) lock-less parallel collision detection algorithm to determine which pebbles are likely to be in contact. The new collision detection algorithm improves on the traditional non-parallel O(n log(n)) collision detection algorithm. These features combine to form a fast parallel pebble motion simulation. The PEBBLES code provides new capabilities for understanding and optimizing PBRs. The PEBBLES code has provided the pebble motion data required to calculate the motion of pebbles during a simulated earthquake. The PEBBLES code provides the ability to

  20. Computational prediction of dust production in graphite moderated pebble bed reactors

    Science.gov (United States)

    Rostamian, Maziar

    The scope of the work reported here, which is the computational study of graphite wear behavior, supports the Nuclear Engineering University Programs project "Experimental Study and Computational Simulations of Key Pebble Bed Thermomechanics Issues for Design and Safety" funded by the US Department of Energy. In this work, modeling and simulating the contact mechanics, as anticipated in a PBR configuration, is carried out for the purpose of assessing the amount of dust generated during a full power operation year of a PBR. A methodology that encompasses finite element analysis (FEA) and micromechanics of wear is developed to address the issue of dust production and its quantification. Particularly, the phenomenon of wear and change of its rate with sliding length is the main focus of this dissertation. This work studies the wear properties of graphite by simulating pebble motion and interactions of a specific type of nuclear grade graphite, IG-11. This study consists of two perspectives: macroscale stress analysis and microscale analysis of wear mechanisms. The first is a set of FEA simulations considering pebble-pebble frictional contact. In these simulations, the mass of generated graphite particulates due to frictional contact is calculated by incorporating FEA results into Archard's equation, which is a linear correlation between wear mass and wear length. However, the experimental data by Johnson, University of Idaho, revealed that the wear rate of graphite decreases with sliding length. This is because the surfaces of the graphite pebbles become smoother over time, which results in a gradual decrease in wear rate. In order to address the change in wear rate, a more detailed analysis of wear mechanisms at room temperature is presented. In this microscale study, the wear behavior of graphite at the asperity level is studied by simulating the contact between asperities of facing surfaces. By introducing the effect of asperity removal on wear rate, a nonlinear

  1. A study for fuel reloading strategy in pebble bed core

    International Nuclear Information System (INIS)

    Kim, Hong Chul

    2012-02-01

    A fuel reloading analysis system for pebble bed reactor was developed by using a Monte Carlo code. The kinematic model was modified to improve the accuracy of the pebble velocity profile and to develop the model so that the diffusion coefficient is not changed by the geometry of the core. In addition, the point kernel method was employed to solve an equation derived in this study. Then, the analysis system for the pebble bed reactor was developed to accommodate the double heterogeneity, pebble velocity, and pebble refueling features using the MCNPX Monte Carlo code. The batch-tracking method was employed to simulate the movement of the pebbles and an automation system was written in the C programming language to implement it. The proposed analysis system can be utilized to verify new core analysis codes, deep-burn studies, various sensitivity studies, and other analysis tools available for the application of new fuel reloading strategies. It is noted that the proposed algorithm for the optimum fuel reloading pattern differs from other optimization methods using sensitivity analysis. In this algorithm, the reloading strategy, including the loading of fresh fuel and the reloading positions of the fresh and reloaded fuels, is determined by the interrelations of the criticality, the nuclear material inventories in the extracted fuel, and the power density. The devised algorithm was applied to the PBMR and NHDD-PBR200. The results show that the proposed algorithm can apply to satisfy the nuclear characteristics such as the criticality or power density since the pebble bed core has the characteristics that the fuels are reloaded every day

  2. Optimized Core Design and Fuel Management of a Pebble-Bed Type Nuclear Reactor

    International Nuclear Information System (INIS)

    Boer, Brian

    2007-01-01

    The Very High Temperature Reactor (VHTR) has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service in the second half of the 21st century. The VHTR is characterized by a high plant efficiency and a high fuel discharge burnup level. More specifically, the (pebble-bed type) High Temperature Reactor (HTR) is known for its inherently safe characteristics, coming from a negative temperature reactivity feedback, a low power density and a large thermal inertia of the core. The core of a pebble-bed reactor consists of graphite spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. The pebbles contain thousands of fuel particles, which are coated with several pyrocarbon and silicon carbon layers that are designed to contain the fission products that are formed during operation of the reactor. The inherent safety concept has been demonstrated in small pebble-bed reactors in practice, but an increase in the reactor size and power is required for cost-effective power production. An increase of the power density in order to increase the helium coolant outlet temperature is attractive with regard to the efficiency and possible process heat applications. However, this increase leads in general to higher fuel temperatures, which could lead to a consequent increase of the fuel coating failure probability. This thesis deals with the pebble-bed type VHTR that aims at an increased coolant outlet temperature of 1000 degrees C and beyond. For the simulation of the neutronic and thermal-hydraulic behavior of the reactor the DALTON-THERMIX coupled code system has been developed and has been validated against experiments performed in the AVR and HTR-10 reactors. An analysis of the 400 MWth Pebble Bed Modular Reactor (PBMR) design shows that the inherent safety concept that has been demonstrated in practice in the smaller AVR and HTR-10

  3. Effect of fuel particles' size variations on multiplication factor in pebble-bed nuclear reactor

    International Nuclear Information System (INIS)

    Snoj, L.; Ravnik, M.

    2005-01-01

    The pebble-bed reactor (Pbr) spherical fuel element consists of two radial zones: the inner zone, in which the fissile material in form of the so-called TRISO particles is uniformly dispersed in graphite matrix and the outer zone, a shell of pure graphite. A TRISO particle is composed of a fissile kernel (UO 2 ) and several layers of carbon composites. The effect of TRISO particles' size variations and distance between them on PBR multiplication factor is studied using MCNP code. Fuel element is modelled in approximation of a cubical unit cell with periodic boundary condition. The multiplication factor of the fuel element depends on the size of the TRISO particles due to resonance self-shielding effect and on the inter-particle distance due to inter-kernel shadowing. (author)

  4. Irradiation creep performance of graphite relevant for pebble bed HTRs

    International Nuclear Information System (INIS)

    Kleist, G.; O'Connor, M.F.

    1980-01-01

    Irradiation - induced creep in the core reflector component graphite of high temperature reactors is of primary importance to the core designer since it provides a mechanism for the relief of internal stresses arising from differential Wigner shrinkage and thermal expansion. The experimental determination of the extent of this creep for conditions relevant to the reactor is thus imperative

  5. Three-Dimensional Analysis of the Hot-Spot Fuel Temperature in Pebble Bed and Prismatic Modular Reactors

    International Nuclear Information System (INIS)

    In, W. K.; Lee, S. W.; Lim, H. S.; Lee, W. J.

    2006-01-01

    High temperature gas-cooled reactors(HTGR) have been reviewed 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 a PBR and a PMR. The objective of this study is to predict the hot-spot fuel temperature distributions in a PBR and a PMR at a steady state. The computational fluid dynamics(CFD) code, CFX-10 is used to perform the three-dimensional analysis. The latest design data was used here based on the reference reactor designs, PBMR400 and GTMHR60

  6. On-line interrogation of pebble bed reactor fuel using passive gamma-ray spectrometry

    Science.gov (United States)

    Chen, Jianwei

    The Pebble Bed Reactor (PBR) is a helium-cooled, graphite-moderated high temperature nuclear power reactor. In addition to its inherently safe design, a unique feature of this reactor is its multipass fuel cycle in which graphite fuel pebbles (of varying enrichment) are randomly loaded and continuously circulated through the core until they reach their prescribed end-of-life burnup limit (˜80,000--100,000 MWD/MTU). Unlike the situation with conventional light water reactors (LWRs), depending solely on computational methods to perform in-core fuel management will be highly inaccurate. As a result, an on-line measurement approach becomes the only accurate method to assess whether a particular pebble has reached its end-of-life burnup limit. In this work, an investigation was performed to assess the feasibility of passive gamma-ray spectrometry assay as an approach for on-line interrogation of PBR fuel for the simultaneous determination of burnup and enrichment on a pebble-by-pebble basis. Due to the unavailability of irradiated or fresh pebbles, Monte Carlo simulations were used to study the gamma-ray spectra of the PBR fuel at various levels of burnup. A pebble depletion calculation was performed using the ORIGEN code, which yielded the gamma-ray source term that was introduced into the input of an MCNP simulation. The MCNP simulation assumed the use of a high-purity coaxial germanium detector. Due to the lack of one-group high temperature reactor cross sections for ORIGEN, a heterogeneous MCNP model was developed to describe a typical PBR core. Subsequently, the code MONTEBURNS was used to couple the MCNP model and ORIGEN. This approach allowed the development of the burnup-dependent, one-group spectral-averaged PBR cross sections to be used in the ORIGEN pebble depletion calculation. Based on the above studies, a relative approach for performing the measurements was established. The approach is based on using the relative activities of Np-239/I-132 in combination

  7. Pebble bed modular reactor fuel enrichment discrimination using delayed neutrons - HTR2008-58133

    International Nuclear Information System (INIS)

    Skoda, R.; Rataj, J.; Uhera, J.

    2008-01-01

    The Pebble Bed Modular Reactor (PBMR) is a helium-cooled, graphite-moderated high temperature nuclear power reactor which utilise fuel in form of spheres that are randomly loaded and continuously circulated through the core until they reach their prescribed end-of-life burn-up limit. When the reactor is started up for the first time, the lower-enriched start-up fuel is used, mixed with graphite spheres, to bring the core to criticality. As the core criticality is established and the start-up fuel is burned-in, the graphite spheres are progressively removed and replaced with more start-up fuel. Once it becomes necessary for maintaining power output, the higher enriched equilibrium fuel is introduced to the reactor and the start-up fuel is removed. During the initial run of the reactor it is important to discriminate between the irradiated startup fuel and the irradiated equilibrium fuel to ensure that only the equilibrium fuel is returned to the reactor. There is therefore a need for an on-line enrichment discrimination device that can discriminate between irradiated start-up fuel spheres and irradiated equilibrium fuel spheres. The device must also not be confused by the presence of any remaining graphite spheres. Due to it's on-line nature the device must accomplish the discrimination within tight time limits. Theoretical calculations and experiments show that Fuel Enrichment Discrimination based on delayed neutrons detection is possible. The paper presents calculations and experiments showing viability of the method. (authors)

  8. Pebble bed reactor fuel cycle optimization using particle swarm algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Tavron, Barak, E-mail: btavron@bgu.ac.il [Planning, Development and Technology Division, Israel Electric Corporation Ltd., P.O. Box 10, Haifa 31000 (Israel); Shwageraus, Eugene, E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, Trumpington Street, Cambridge CB2 1PZ (United Kingdom)

    2016-10-15

    Highlights: • Particle swarm method has been developed for fuel cycle optimization of PBR reactor. • Results show uranium utilization low sensitivity to fuel and core design parameters. • Multi-zone fuel loading pattern leads to a small improvement in uranium utilization. • Thorium mixes with highly enriched uranium yields the best uranium utilization. - Abstract: Pebble bed reactors (PBR) features, such as robust thermo-mechanical fuel design and on-line continuous fueling, facilitate wide range of fuel cycle alternatives. A range off fuel pebble types, containing different amounts of fertile or fissile fuel material, may be loaded into the reactor core. Several fuel loading zones may be used since radial mixing of the pebbles was shown to be limited. This radial separation suggests the possibility to implement the “seed-blanket” concept for the utilization of fertile fuels such as thorium, and for enhancing reactor fuel utilization. In this study, the particle-swarm meta-heuristic evolutionary optimization method (PSO) has been used to find optimal fuel cycle design which yields the highest natural uranium utilization. The PSO method is known for solving efficiently complex problems with non-linear objective function, continuous or discrete parameters and complex constrains. The VSOP system of codes has been used for PBR fuel utilization calculations and MATLAB script has been used to implement the PSO algorithm. Optimization of PBR natural uranium utilization (NUU) has been carried out for 3000 MWth High Temperature Reactor design (HTR) operating on the Once Trough Then Out (OTTO) fuel management scheme, and for 400 MWth Pebble Bed Modular Reactor (PBMR) operating on the multi-pass (MEDUL) fuel management scheme. Results showed only a modest improvement in the NUU (<5%) over reference designs. Investigation of thorium fuel cases showed that the use of HEU in combination with thorium results in the most favorable reactor performance in terms of

  9. Pebble bed reactor fuel cycle optimization using particle swarm algorithm

    International Nuclear Information System (INIS)

    Tavron, Barak; Shwageraus, Eugene

    2016-01-01

    Highlights: • Particle swarm method has been developed for fuel cycle optimization of PBR reactor. • Results show uranium utilization low sensitivity to fuel and core design parameters. • Multi-zone fuel loading pattern leads to a small improvement in uranium utilization. • Thorium mixes with highly enriched uranium yields the best uranium utilization. - Abstract: Pebble bed reactors (PBR) features, such as robust thermo-mechanical fuel design and on-line continuous fueling, facilitate wide range of fuel cycle alternatives. A range off fuel pebble types, containing different amounts of fertile or fissile fuel material, may be loaded into the reactor core. Several fuel loading zones may be used since radial mixing of the pebbles was shown to be limited. This radial separation suggests the possibility to implement the “seed-blanket” concept for the utilization of fertile fuels such as thorium, and for enhancing reactor fuel utilization. In this study, the particle-swarm meta-heuristic evolutionary optimization method (PSO) has been used to find optimal fuel cycle design which yields the highest natural uranium utilization. The PSO method is known for solving efficiently complex problems with non-linear objective function, continuous or discrete parameters and complex constrains. The VSOP system of codes has been used for PBR fuel utilization calculations and MATLAB script has been used to implement the PSO algorithm. Optimization of PBR natural uranium utilization (NUU) has been carried out for 3000 MWth High Temperature Reactor design (HTR) operating on the Once Trough Then Out (OTTO) fuel management scheme, and for 400 MWth Pebble Bed Modular Reactor (PBMR) operating on the multi-pass (MEDUL) fuel management scheme. Results showed only a modest improvement in the NUU (<5%) over reference designs. Investigation of thorium fuel cases showed that the use of HEU in combination with thorium results in the most favorable reactor performance in terms of

  10. Nonproliferation and safeguard considerations: Pebble Bed reactor fuel cycle evaluation

    International Nuclear Information System (INIS)

    1978-09-01

    Nuclear fuel cycles were evaluated for the Pebble Bed Gas Cooled Reactor under development in the Federal Republic of Germany. The basic fuel cycle specified for the HTR-K and PNP is well qualified and will meet the requirements of these reactors. Twenty alternate fuel cycles are described, including high-conversion cycles, net-breeding cycles, and proliferation-resistant cycles. High-conversion cycles, which have a high probability of being successfully developed, promise a significant improvement in resource utilization. Proliferation-resistant cycles, also with a high probability of successful development, conpare very favorably with those for other types of reactors. Most of the advanced cycles could be adapted to first-generation pebble bed reactors with no significant modifications

  11. Analytical calculation of the fuel temperature reactivity coefficient for pebble bed and prismatic high temperature reactors for plutonium and uranium-thorium fuels

    International Nuclear Information System (INIS)

    Talamo, Alberto

    2007-01-01

    We analytically evaluated the fuel coefficient of temperature both for pebble bed and prismatic high temperature reactors when they utilize as fuel plutonium and minor actinides from light water reactors spent fuel or a mixture of 50% uranium, enriched 20% in 235 U, and 50% thorium. In both cores the calculation involves the evaluation of the resonances integrals of the high absorbers fuel nuclides 240 Pu, 238 U and 232 Th and it requires the esteem of the Dancoff-Ginsburg factor for a pebble bed or prismatic core. The Dancoff-Ginsburg factor represents the only discriminating parameter in the results for the two different reactors types; in fact, both the pebble bed and the prismatic reactors share the same the pseudo-cross-section describing an infinite medium made of graphite filled by TRISO particles. We considered only the resolved resonances with a statistical spin factor equal to one and we took into account 267, 72, 212 resonances in the range 1.057-5692, 6.674-14485, 21.78-3472 eV for 240 Pu, 238 U and 232 Th, respectively, for investigating the influence on the fuel temperature reactivity coefficient of the variation of the TRISO kernel radius and TRISO particles packing fraction from 100, 200 to 300 μm and from 10% to 50%, respectively. Finally, in the pebble bed core, we varied the radius of the pebble for setting a fuel temperature reactivity coefficient similar to the one of a prismatic core

  12. Pebble-bed pebble motion: Simulation and Applications

    Energy Technology Data Exchange (ETDEWEB)

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2011-11-01

    Pebble bed reactors (PBR) have moving graphite fuel pebbles. This unique feature provides advantages, but also means that simulation of the reactor requires understanding the typical motion and location of the granular flow of pebbles. This report presents a method for simulation of motion of the pebbles in a PBR. A new mechanical motion simulator, PEBBLES, efficiently simulates the key elements of motion of the pebbles in a PBR. This model simulates gravitational force and contact forces including kinetic and true static friction. It's used for a variety of tasks including simulation of the effect of earthquakes on a PBR, calculation of packing fractions, Dancoff factors, pebble wear and the pebble force on the walls. The simulator includes a new differential static friction model for the varied geometries of PBRs. A new static friction benchmark was devised via analytically solving the mechanics equations to determine the minimum pebble-to-pebble friction and pebble-to-surface friction for a five pebble pyramid. This pyramid check as well as a comparison to the Janssen formula was used to test the new static friction equations. Because larger pebble bed simulations involve hundreds of thousands of pebbles and long periods of time, the PEBBLES code has been parallelized. PEBBLES runs on shared memory architectures and distributed memory architectures. For the shared memory architecture, the code uses a new O(n) lock-less parallel collision detection algorithm to determine which pebbles are likely to be in contact. The new collision detection algorithm improves on the traditional non-parallel O(n log(n)) collision detection algorithm. These features combine to form a fast parallel pebble motion simulation. The PEBBLES code provides new capabilities for understanding and optimizing PBRs. The PEBBLES code has provided the pebble motion data required to calculate the motion of pebbles during a simulated earthquake. The PEBBLES code provides the ability to

  13. Pebble-bed pebble motion: Simulation and Applications

    International Nuclear Information System (INIS)

    Cogliati, Joshua J.; Ougouag, Abderrafi M.

    2011-01-01

    Pebble bed reactors (PBR) have moving graphite fuel pebbles. This unique feature provides advantages, but also means that simulation of the reactor requires understanding the typical motion and location of the granular flow of pebbles. This report presents a method for simulation of motion of the pebbles in a PBR. A new mechanical motion simulator, PEBBLES, efficiently simulates the key elements of motion of the pebbles in a PBR. This model simulates gravitational force and contact forces including kinetic and true static friction. It's used for a variety of tasks including simulation of the effect of earthquakes on a PBR, calculation of packing fractions, Dancoff factors, pebble wear and the pebble force on the walls. The simulator includes a new differential static friction model for the varied geometries of PBRs. A new static friction benchmark was devised via analytically solving the mechanics equations to determine the minimum pebble-to-pebble friction and pebble-to-surface friction for a five pebble pyramid. This pyramid check as well as a comparison to the Janssen formula was used to test the new static friction equations. Because larger pebble bed simulations involve hundreds of thousands of pebbles and long periods of time, the PEBBLES code has been parallelized. PEBBLES runs on shared memory architectures and distributed memory architectures. For the shared memory architecture, the code uses a new O(n) lock-less parallel collision detection algorithm to determine which pebbles are likely to be in contact. The new collision detection algorithm improves on the traditional non-parallel O(n log(n)) collision detection algorithm. These features combine to form a fast parallel pebble motion simulation. The PEBBLES code provides new capabilities for understanding and optimizing PBRs. The PEBBLES code has provided the pebble motion data required to calculate the motion of pebbles during a simulated earthquake. The PEBBLES code provides the ability to determine

  14. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part I: Pebble Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brian Boer; Abderrafi M. Ougouag

    2011-03-01

    The Deep-Burn (DB) concept [ ] focuses on the destruction of transuranic nuclides from used light water reactor (LWR) fuel. These transuranic nuclides are incorporated into tri-isotopic (TRISO) coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400) [ ]. Although it has been shown in the previous Fiscal Year (FY) (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking, and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239Pu, 240Pu, and 241Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a standard, UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. Regarding the coated particle performance, the FY 2009 investigations showed that no

  15. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part I: Pebble Bed Reactors

    International Nuclear Information System (INIS)

    Boer, Brian; Ougouag, Abderrafi M.

    2011-01-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor (LWR) fuel. These transuranic nuclides are incorporated into tri-isotopic (TRISO) coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (FY) (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking, and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239Pu, 240Pu, and 241Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a standard, UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. Regarding the coated particle performance, the FY 2009 investigations showed that no significant

  16. An Analysis of Fuel Region to Region Dancoff Factor with the Random Mixture Effects of Moderator and Fuel Pebbles

    International Nuclear Information System (INIS)

    Kim, Song Hyun; Kim, Hong Chul; Kim, Jong Kyung; Noh, Jae Man

    2009-01-01

    Dancoff factor is an entering probability of the neutron escaped from specific fuel kernel to another one without the interaction with moderators. In order to analytically evaluate Dancoff factor considering double-heterogeneous effect, inter-pebble and intra-pebble Dancoff factors should be calculated, respectively. Intra-pebble Dancoff factor related with the fuel kernels in one pebble was analyzed in the past study. The fuel and moderator pebbles are randomly located in the pebble-type reactor. For the evaluation of inter-pebble Dancoff factor, a repetition of simple pebble structure is commonly assumed to simulate the complex geometry of pebble-type reactor. The evaluation using these structures can be underestimated because of the shadowing effects generated from the repetition of simple pebble structure. Fuel region to region Dancoff factor (FRDF) was defined as an entering probability of the neutron escaped from a specific fuel region to another one without any collision with moderator for a preliminary evaluation of inter-pebble Dancoff factor. To solve the underestimation problem of FRDF from the shadow effect, the specific pebble was assumed and FRDF was evaluated with the approximation method proposed in this study

  17. Power Peaking Effect of OTTO Fuel Scheme Pebble Bed Reactor

    Science.gov (United States)

    Setiadipura, T.; Suwoto; Zuhair; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    Pebble Bed Reactor (PBR) type of Hight Temperature Gas-cooled Reactor (HTGR) is a very interesting nuclear reactor design to fulfill the growing electricity and heat demand with a superior passive safety features. Effort to introduce the PBR design to the market can be strengthen by simplifying its system with the Once-through-then-out (OTTO) cycle PBR in which the pebble fuel only pass the core once. Important challenge in the OTTO fuel scheme is the power peaking effect which limit the maximum nominal power or burnup of the design. Parametric survey is perform in this study to investigate the contribution of different design parameters to power peaking effect of OTTO cycle PBR. PEBBED code is utilized in this study to perform the equilibrium PBR core analysis for different design parameter and fuel scheme. The parameters include its core diameter, height-per-diameter (H/D), power density, and core nominal power. Results of this study show that diameter and H/D effectsare stronger compare to the power density and nominal core power. Results of this study might become an importance guidance for design optimization of OTTO fuel scheme PBR.

  18. Year One Summary of X-energy Pebble Fuel Development at ORNL

    Energy Technology Data Exchange (ETDEWEB)

    Helmreich, Grant W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McMurray, Jake W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunt, Rodney D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jolly, Brian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Trammell, Michael P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Daniel R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Blamer, Brandon J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Reif, Tyler J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kim, Howard T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-06-01

    The Advanced Reactor Concepts X-energy (ARC-Xe) Pebble Fuel Development project at Oak Ridge National Laboratory (ORNL) has successfully completed its first year, having made excellent progress in accomplishing programmatic objectives. The primary focus of research at ORNL in support of X-energy has been the training of X-energy fuel fabrication engineers and the establishment of US pebble fuel production capabilities able to supply the Xe-100 pebble-bed reactor. These efforts have been strongly supported by particle fuel fabrication and characterization expertise present at ORNL from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program.

  19. Analytical calculation of the fuel temperature reactivity coefficient for pebble bed and prismatic high temperature reactors for plutonium and uranium-thorium fuels

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto [Department of Nuclear and Reactor Physics, Royal Institute of Technology - KTH, Roslagstullsbacken 21, S-10691 Stockholm (Sweden)]. E-mail: alby@anl.gov

    2007-01-15

    We analytically evaluated the fuel coefficient of temperature both for pebble bed and prismatic high temperature reactors when they utilize as fuel plutonium and minor actinides from light water reactors spent fuel or a mixture of 50% uranium, enriched 20% in {sup 235}U, and 50% thorium. In both cores the calculation involves the evaluation of the resonances integrals of the high absorbers fuel nuclides {sup 240}Pu, {sup 238}U and {sup 232}Th and it requires the esteem of the Dancoff-Ginsburg factor for a pebble bed or prismatic core. The Dancoff-Ginsburg factor represents the only discriminating parameter in the results for the two different reactors types; in fact, both the pebble bed and the prismatic reactors share the same the pseudo-cross-section describing an infinite medium made of graphite filled by TRISO particles. We considered only the resolved resonances with a statistical spin factor equal to one and we took into account 267, 72, 212 resonances in the range 1.057-5692, 6.674-14485, 21.78-3472 eV for {sup 240}Pu, {sup 238}U and {sup 232}Th, respectively, for investigating the influence on the fuel temperature reactivity coefficient of the variation of the TRISO kernel radius and TRISO particles packing fraction from 100, 200 to 300 {mu}m and from 10% to 50%, respectively. Finally, in the pebble bed core, we varied the radius of the pebble for setting a fuel temperature reactivity coefficient similar to the one of a prismatic core.

  20. Improvement of burnup analysis for pebble bed reactors with an accumulative fuel loading scheme

    International Nuclear Information System (INIS)

    Simanullang, Irwan Liapto; Obara, Toru

    2015-01-01

    Given the limitations of natural uranium resources, innovative nuclear power plant concepts that increase the efficiency of nuclear fuel utilization are needed. The Pebble Bed Reactor (PBR) shows some potential to achieve high efficiency in natural uranium utilization. To simplify the PBR concept, PBR with an accumulation fuel loading scheme was introduced and the Fuel Handling System (FHS) removed. In this concept, the pebble balls are added little by little into the reactor core until the pebble balls reach the top of the reactor core, and all pebble balls are discharged from the core at the end of the operation period. A code based on the MVP/MVP-BURN method has been developed to perform an analysis of a PBR with the accumulative fuel loading scheme. The optimum fuel composition was found using the code for high burnup performance. Previous efforts provided several motivations to improve the burnup performance: First, some errors in the input code were corrected. This correction, and an overall simplification of the input code, was implemented for easier analysis of a PBR with the accumulative fuel loading scheme. Second, the optimum fuel design had been obtained in the infinite geometry. To improve the optimum fuel composition, a parametric survey was obtained by varying the amount of Heavy Metal (HM) uranium per pebble and the degree of uranium enrichment. Moreover, an entire analysis of the parametric survey was obtained in the finite geometry. The results show that improvements in the fuel composition can lead to more accurate analysis with the code. (author)

  1. Comparison of Material Behavior of Matrix Graphite for HTGR Fuel Elements upon Irradiation: A literature Survey

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The fuel elements for the HTGRs (i.e., spherical fuel element in pebble-bed type core design and fuel compact in prismatic core design) consists of coated fuel particles dispersed and bonded in a closely packed array within a carbonaceous matrix. This matrix is generally made by mixing fully graphitized natural and needle- or pitchcoke originated powders admixed with a binder material (pitch or phenolic resin), The resulting resinated graphite powder mixture, when compacted, may influence a number of material properties as well as its behavior under neutron irradiation during reactor operation. In the fabrication routes of these two different fuel element forms, different consolidation methods are employed; a quasi-isostatic pressing method is generally adopted to make pebbles while fuel compacts are fabricated by uni-axial pressing mode. The result showed that the hardness values obtained from the two directions showed an anisotropic behavior: The values obtained from the perpendicular section showed much higher micro hardness (176.6±10.5MPa in average) than from the parallel section ((125.6±MPa in average). This anisotropic behavior was concluded to be related to the microstructure of the matrix graphite. This may imply that the uni-axial pressing method to make compacts influence the microstructure of the matrix and hence the material properties of the matrix graphite.

  2. Numerical Simulation of Particle Flow Motion in a Two-Dimensional Modular Pebble-Bed Reactor with Discrete Element Method

    Directory of Open Access Journals (Sweden)

    Guodong Liu

    2013-01-01

    Full Text Available Modular pebble-bed nuclear reactor (MPBNR technology is promising due to its attractive features such as high fuel performance and inherent safety. Particle motion of fuel and graphite pebbles is highly associated with the performance of pebbled-bed modular nuclear reactor. To understand the mechanism of pebble’s motion in the reactor, we numerically studied the influence of number ratio of fuel and graphite pebbles, funnel angle of the reactor, height of guide ring on the distribution of pebble position, and velocity by means of discrete element method (DEM in a two-dimensional MPBNR. Velocity distributions at different areas of the reactor as well as mixing characteristics of fuel and graphite pebbles were investigated. Both fuel and graphite pebbles moved downward, and a uniform motion was formed in the column zone, while pebbles motion in the cone zone was accelerated due to the decrease of the cross sectional flow area. The number ratio of fuel and graphite pebbles and the height of guide ring had a minor influence on the velocity distribution of pebbles, while the variation of funnel angle had an obvious impact on the velocity distribution. Simulated results agreed well with the work in the literature.

  3. Pebble Bed Reactor Dust Production Model

    Energy Technology Data Exchange (ETDEWEB)

    Abderrafi M. Ougouag; Joshua J. Cogliati

    2008-09-01

    The operation of pebble bed reactors, including fuel circulation, can generate graphite dust, which in turn could be a concern for internal components; and to the near field in the remote event of a break in the coolant circuits. The design of the reactor system must, therefore, take the dust into account and the operation must include contingencies for dust removal and for mitigation of potential releases. Such planning requires a proper assessment of the dust inventory. This paper presents a predictive model of dust generation in an operating pebble bed with recirculating fuel. In this preliminary work the production model is based on the use of the assumption of proportionality between the dust production and the normal force and distance traveled. The model developed in this work uses the slip distances and the inter-pebble forces computed by the authors’ PEBBLES. The code, based on the discrete element method, simulates the relevant static and kinetic friction interactions between the pebbles as well as the recirculation of the pebbles through the reactor vessel. The interaction between pebbles and walls of the reactor vat is treated using the same approach. The amount of dust produced is proportional to the wear coefficient for adhesive wear (taken from literature) and to the slip volume, the product of the contact area and the slip distance. The paper will compare the predicted volume with the measured production rates. The simulation tallies the dust production based on the location of creation. Two peak production zones from intra pebble forces are predicted within the bed. The first zone is located near the pebble inlet chute due to the speed of the dropping pebbles. The second peak zone occurs lower in the reactor with increased pebble contact force due to the weight of supported pebbles. This paper presents the first use of a Discrete Element Method simulation of pebble bed dust production.

  4. Pebble Bed Reactor Dust Production Model

    International Nuclear Information System (INIS)

    Abderrafi M. Ougouag; Joshua J. Cogliati

    2008-01-01

    The operation of pebble bed reactors, including fuel circulation, can generate graphite dust, which in turn could be a concern for internal components; and to the near field in the remote event of a break in the coolant circuits. The design of the reactor system must, therefore, take the dust into account and the operation must include contingencies for dust removal and for mitigation of potential releases. Such planning requires a proper assessment of the dust inventory. This paper presents a predictive model of dust generation in an operating pebble bed with recirculating fuel. In this preliminary work the production model is based on the use of the assumption of proportionality between the dust production and the normal force and distance traveled. The model developed in this work uses the slip distances and the inter-pebble forces computed by the authors PEBBLES. The code, based on the discrete element method, simulates the relevant static and kinetic friction interactions between the pebbles as well as the recirculation of the pebbles through the reactor vessel. The interaction between pebbles and walls of the reactor vat is treated using the same approach. The amount of dust produced is proportional to the wear coefficient for adhesive wear (taken from literature) and to the slip volume, the product of the contact area and the slip distance. The paper will compare the predicted volume with the measured production rates. The simulation tallies the dust production based on the location of creation. Two peak production zones from intra pebble forces are predicted within the bed. The first zone is located near the pebble inlet chute due to the speed of the dropping pebbles. The second peak zone occurs lower in the reactor with increased pebble contact force due to the weight of supported pebbles. This paper presents the first use of a Discrete Element Method simulation of pebble bed dust production

  5. An Experiment on the Carbonization of Fuel Compact Matrix Graphite for HTGR

    International Nuclear Information System (INIS)

    Lee, Young Woo; Kim, Joo Hyoung; Cho, Moon Sung

    2012-01-01

    The fuel element for HTGR is manufactured by mixing coated fuel particles with matrix graphite powder and forming into either pebble type or cylindrical type compacts depending on their use in different HTGR cores. The coated fuel particle, the so-called TRISO particle, consists of 500-μm spherical UO 2 particles coated with the low density buffer Pyrolytic Carbon (PyC) layer, the inner and outer high density PyC layer and SiC layer sandwiched between the two inner and outer PyC layers. The coated TRISO particles are mixed with a properly prepared matrix graphite powder, pressed into a spherical shape or a cylindrical compact, and finally heat-treated at about 1800 .deg. C. These fuel elements can have different sizes and forms of compact. The basic steps for manufacturing a fuel element include preparation of graphite matrix powder, over coating the fuel particles, mixing the fuel particles with a matrix powder, carbonizing green compact, and the final high-temperature heat treatment of the carbonized fuel compact. The carbonization is a process step where the binder that is incorporated during the matrix graphite powder preparation step is evaporated and the residue of the binder is carbonized during the heat treatment at about 1073 K, In order to develop a fuel compact fabrication technology, and for fuel matrix graphite to meet the required material properties, it is of extreme importance to investigate the relationship among the process parameters of the matrix graphite powder preparation, fabrication parameters of fuel element green compact and the carbonization condition, which has a strong influence on further steps and the material properties of fuel element. In this work, the carbonization behavior of green compact samples prepared from the matrix graphite powder mixtures with different binder materials was investigated in order to elucidate the behavior of binders during the carbonization heat treatment by analyzing the change in weight, density and its

  6. Burnup performance of OTTO cycle pebble bed reactors with ROX fuel

    International Nuclear Information System (INIS)

    Ho, Hai Quan; Obara, Toru

    2015-01-01

    Highlights: • A 300 MW t Small Pebble Bed Reactor with Rock-like oxide fuel is proposed. • Using ROX fuel can achieve high discharged burnup of spent fuel. • High geological stability can be expected in direct disposal of the spent ROX fuel. • The Pebble Bed Reactor with ROX fuel can be critical at steady state operation. • All the reactor designs have a negative temperature coefficient. - Abstract: A pebble bed high-temperature gas-cooled reactor (PBR) with rock-like oxide (ROX) fuel was designed to achieve high discharged burnup and improve the integrity of the spent fuel in geological disposal. The MCPBR code with a JENDL-4.0 library, which developed the analysis of the Once-Through-Then-Out (OTTO) cycle in PBR, was used to perform the criticality and burnup analysis. Burnup calculations for eight cases were carried out for both ROX fuel and a UO 2 fuel reactor with different heavy-metal loading conditions. The effective multiplication factor of all cases approximately equalled unity in the equilibrium condition. The ROX fuel reactor showed lower FIFA than the UO 2 fuel reactor at the same heavy-metal loading, about 5–15%. However, the power peaking factor and maximum power per fuel ball in the ROX fuel core were lower than that of UO 2 fuel core. This effect makes it possible to compensate for the lower-FIFA disadvantage in a ROX fuel core. All reactor designs had a negative temperature coefficient that is needed for the passive safety features of a pebble bed reactor

  7. Interim report on core physics and fuel cycle analysis of the pebble bed reactor power plant concept

    International Nuclear Information System (INIS)

    Vondy, D.R.

    1977-12-01

    Calculations were made to predict the performance of a pebble bed reactor operated in a mode to produce fissile fuel (high conversion or breeding). Both a one pebble design and a design involving large primary feed pebbles and small fertile pebbles were considered. A relatively short residence time of the primary pebbles loaded with 233 U fuel was found to be necessary to achieve a high breeding ratio, but this leads to relatively high fuel costs. A high fissile inventory is associated with a low C/Th ratio and a high thorium loading, causing the doubling time to be long, even though the breeding ratio is high, and the fuel cost of electrical product to be high. Production of 233 U fuel from 235 U feed was studied and performances of the converter and breeder reactor concepts were examined varying the key parameters

  8. Random detailed model for probabilistic neutronic calculation in pebble bed Very High Temperature Reactors

    International Nuclear Information System (INIS)

    Perez Curbelo, J.; Rosales, J.; Garcia, L.; Garcia, C.; Brayner, C.

    2013-01-01

    The pebble bed nuclear reactor is one of the main candidates for the next generation of nuclear power plants. In pebble bed type HTRs, the fuel is contained within graphite pebbles in the form of TRISO particles, which form a randomly packed bed inside a graphite-walled cylindrical cavity. Pebble bed reactors (PBR) offer the opportunity to meet the sustainability requirements, such as nuclear safety, economic competitiveness, proliferation resistance and a minimal production of radioactive waste. In order to simulate PBRs correctly, the double heterogeneity of the system must be considered. It consists on randomly located pebbles into the core and TRISO particles into the fuel pebbles. These features are often neglected due to the difficulty to model with MCPN code. The main reason is that there is a limited number of cells and surfaces to be defined. In this study, a computational tool which allows getting a new geometrical model of fuel pebbles for neutronic calculations with MCNPX code, was developed. The heterogeneity of system is considered, and also the randomly located TRISO particles inside the pebble. Four proposed fuel pebble models were compared regarding their effective multiplication factor and energy liberation profiles. Such models are: Homogeneous Pebble, Five Zone Homogeneous Pebble, Detailed Geometry, and Randomly Detailed Geometry. (Author)

  9. Computational Investigation of On-Line Interrogation of Pebble Bed Reactor Fuel

    Science.gov (United States)

    Hawari, A. I.; Chen, Jianwei

    2005-10-01

    Pebble bed reactors are characterized by multipass fuel systems in which spherical fuel pebbles are circulated through the core until they reach a proposed burnup limit (80000-100000 MWD/MTU). For such reactors, the fuel is assayed on-line to ensure that the burnup limit is not breached. We considered assaying the fuel using an HPGe detector to perform passive gamma-ray spectrometry of fission products. Since neither fresh nor irradiated fuel is readily available, computer simulations were utilized to identify the radionuclides that can be used as burnup indicators, and to visualize the gamma-ray spectra at various levels of burnup. Specifically, we used the ORIGEN-MONTEBURNS-MCNP code system. This allowed the establishment of the burnup dependent one-group gas reactor cross-sections for the radionuclides of interest. Subsequently, ORIGEN was used to simulate in-core pebble depletion to establish the irradiated pebble isotopics. Finally, the codes MCNP and SYNTH were used to simulate the response of the HPGe gamma-ray spectrometer. The results show that absolute and relative indicators can be used on-line to determine unambiguously the enrichment and burnup on a pebble-by-pebble basis. The activity of Cs-137 or the activity ratio of Co-60/Cs-134 can be combined with the activity ratio of Np-239/I-132 to yield the enrichment and burnup information. To use the relative indicators, a relative efficiency calibration of the gamma-ray spectrometer can be performed using the La-140 gamma lines that are emitted by the irradiated pebble. I-132, Cs-134, Cs-137, La-140, and Np-239 are produced upon the irradiation of the fuel. Co-60 is produced by doping the fuel with a small amount (/spl sim/100 ppm) of Co-59. Using this approach, the uncertainty in burnup determination due to factors such as power history variation, detector efficiency calibration, and counting statistics is expected to remain in the range of /spl plusmn/5% to /spl plusmn/10%.

  10. Status of Research on Pebble Bed HTR Fuel Fabrication Technology in Indonesia

    International Nuclear Information System (INIS)

    Rachmawati, M.; Sarjono; Ridwan; Langenati, R.

    2014-01-01

    Research on pebble bed HTR fuel fabrication is conducted in Indonesia. One of the aims is to build a knowledge base on pebble bed HTR fuel element fabrication technology for fuel procurement. The steps of research strategies are firstly to understand the basic design research of TRISO fuel, properties, and requirements, and secondly to understand the TRISO fuel manufacturing technology, which comprises fabrication and quality control, including its facility. Both steps are adopted from research and experiences of the countries with HTR fuel element fabrication technology. From the knowledge gained in the research, an experimental design of the process and a set of prototype process equipment for fabrication are developed, namely kernels production using external gelation process, TRISO coating of the kernel, and pebble compacting. Experiments using the prototypes have been conducted. Characterization of the kernel product, i.e. diameter, sphericity, density and O/U ratio, shows that the kernel product is still not in compliance with the specification requirements. These are deemed to be caused mainly by the selected vibrating system and the viscosity adjustment. Another major cause is the selected NH3 and air feeding method for both NH3 and air layer in the preparation for spherical droplets of liquid. The FB-CVD TRISO coating of the kernel has been experimented but unsuccessful by using an FB-CVD once‐through continuous coating process. For the pebble compacting, the process is still in the early stage of setting-up compaction equipment. This paper summarizes the current status of research on HTR fuel fabrication technology in Indonesia, the proposed process and its equipment setting-up for improvement of the kernel production. The knowledge and lessons learned gained from the research is useful and can be an assistance in planning for fuel development laboratory facilities procurement, formulating User Requirement Document and Bid Invitation Specification for

  11. Nuclear graphite wear properties and estimation of graphite dust production in HTR-10

    Energy Technology Data Exchange (ETDEWEB)

    Luo, Xiaowei, E-mail: xwluo@tsinghua.edu.cn; Wang, Xiaoxin; Shi, Li; Yu, Xiaoyu; Yu, Suyuan

    2017-04-15

    Highlights: • Graphite dust. • The wear properties of graphite. • Pebble bed. • High Temperature Gas-cooled Reactor. • Fuel element. - Abstract: The issue of the graphite dust has been a research focus for the safety of High Temperature Gas-cooled Reactors (HTGRs), especially for the pebble bed reactors. Most of the graphite dust is produced from the wear of fuel elements during cycling of fuel elements. However, due to the complexity of the motion of the fuel elements in the pebble bed, there is no systematic method developed to predict the amount the graphite dust in a pebble bed reactor. In this paper, the study of the flow of the fuel elements in the pebble bed was carried out. Both theoretical calculation and numerical analysis by Discrete Element Method (DEM) software PFC3D were conducted to obtain the normal forces and sliding distances of the fuel elements in pebble bed. The wearing theory was then integrated with PFC3D to estimate the amount of the graphite dust in a pebble bed reactor, 10 MW High Temperature gas-cooled test Reactor (HTR-10).

  12. Impact on burnup performance of coated particle fuel design in pebble bed reactor with ROX fuel

    International Nuclear Information System (INIS)

    Ho, Hai Quan; Obara, Toru

    2015-01-01

    The pebble bed reactor (PBR), a kind of high-temperature gas-cooled reactor (HTGR), is expected to be among the next generation of nuclear reactors as it has excellent passive safety features, as well as online refueling and high thermal efficiency. Rock-like oxide (ROX) fuel has been studied at the Japan Atomic Energy Agency (JAEA) as a new once-through type fuel concept. Rock-like oxide used as fuel in a PBR can be expected to achieve high burnup and improve chemical stabilities. In the once-through fuel concept, the main challenge is to achieve as high a burnup as possible without failure of the spent fuel. The purpose of this study was to investigate the impact on burnup performance of different coated fuel particle (CFP) designs in a PBR with ROX fuel. In the study, the AGR-1 Coated Particle design and Deep-Burn Coated Particle design were used to make the burnup performance comparison. Criticality and core burnup calculations were performed by MCPBR code using the JENDL-4.0 library. Results at equilibrium showed that the two reactors utilizing AGR-1 Coated Particle and Deep-Burn Coated Particle designs could be critical with almost the same multiplication factor k eff . However, the power peaking factor and maximum power per fuel ball in the AGR-1 coated particle design was lower than that of Deep-Burn coated particle design. The AGR-1 design also showed an advantage in fissions per initial fissile atoms (FIFA); the AGR-1 coated particle design produced a higher FIFA than the Deep-Burn coated particle design. These results suggest that the difference in coated particle fuel design can have an effect on the burnup performance in ROX fuel. (author)

  13. Gas reactor international cooperative program interim report. Pebble bed reactor fuel cycle evaluation

    International Nuclear Information System (INIS)

    1978-09-01

    Nuclear fuel cycles were evaluated for the Pebble Bed Gas Cooled Reactor under development in the Federal Republic of Germany. The basic fuel cycle specified for the HTR-K and PNP is well qualified and will meet the requirements of these reactors. Twenty alternate fuel cycles are described, including high-conversion cycles, net-breeding cycles, and proliferation-resistant cycles. High-conversion cycles, which have a high probability of being successfully developed, promise a significant improvement in resource utilization. Proliferation-resistant cycles, also with a high probability of successful development, compare very favorably with those for other types of reactors. Most of the advanced cycles could be adapted to first-generation pebble bed reactors with no significant modifications

  14. Burnup performance of rock-like oxide (ROX) fuel in small pebble bed reactor with accumulative fuel loading scheme

    International Nuclear Information System (INIS)

    Simanullang, Irwan Liapto; Obara, Toru

    2017-01-01

    Highlights: • Burnup performance using ROX fuel in PBR with accumulative fuel loading scheme was analyzed. • Initial excess reactivity was suppressed by reducing 235 U enrichment in the startup condition. • Negative temperature coefficient was achieved in all condition of PBR with accumulative fuel loading scheme using ROX fuel. • Core lifetime of PBR with accumulative fuel loading scheme using ROX fuel was shorter than with UO 2 fuel. • In PBR with accumulative fuel loading scheme using ROX fuel, achieved discharged burnup can be as high as that for UO 2 fuel. - Abstract: The Japan Atomic Energy Agency (JAEA) has proposed rock-like oxide (ROX) fuel as a new, once-through type fuel concept. Here, burnup performance using ROX fuel was simulated in a pebble bed reactor with an accumulative fuel loading scheme. The MVP-BURN code was used to simulate the burnup calculation. Fuel of 5 g-HM/pebble with 20% 235 U enrichment was selected as the optimum composition. Discharged burnup could reach up to 218 GWd/t, with a core lifetime of about 8.4 years. However, high excess reactivity occurred in the initial condition. Initial fuel enrichment was therefore reduced from 20% to 4.65% to counter the initial excess reactivity. The operation period was reduced by the decrease of initial fuel enrichment, but the maximum discharged burnup was 198 GWd/t. Burnup performance of ROX fuel in this reactor concept was compared with that of UO 2 fuel obtained previously. Discharged burnup for ROX fuel in the PBR with an accumulative fuel loading scheme was as high as UO 2 fuel. Maximum power density could be lowered by introducing ROX fuel compared to UO 2 fuel. However, PBR core lifetime was shorter with ROX fuel than with UO 2 fuel. A negative temperature coefficient was achieved for both UO 2 and ROX fuels throughout the operation period.

  15. CFD simulation of a coolant flow and a heat transfer in a pebble bed reactor - HTR2008-58334

    International Nuclear Information System (INIS)

    In, W. K.; Lee, W. J.; Hassan, Y. A.

    2008-01-01

    This CFD study is to simulate a coolant(gas) flow and heat transfer in a PBR core during a normal operation. This study used a pebble array with direct area contacts among the pebbles which is one of the pebbles arrangements for a detailed simulation of PBR core CFD studies. A CFD model is developed to more adequately represent the pebbles randomly stacked in the PBR core. The CFD predictions showed a large variation of the temperature on the pebble surface as well as in the pebble core. The temperature drop in the outer graphite layer is smaller than that in the pebble-core region. This is because the thermal conductivity of graphite is higher than the fuel (UO, mixture) conductivity in the pebble core. Higher pebble surface temperature is predicted downstream of the pebble contact due to a reverse flow. Multiple vortices are predicted to occur downstream of the spherical pebbles due to a flow separation. The coolant flow structure and fuel temperature in the PBR core appears to largely depend on the in-core distribution of the pebbles. (authors)

  16. Detection system for location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Hongbing, E-mail: liuhb07@mails.tsinghua.edu.cn [Department of Mechanical Engineering, Tsinghua University, Beijing 100084 (China); Key Laboratory for Advanced Materials Processing Technology, Ministry of Education P. R. China, Beijing 100084 (China); Du, Dong, E-mail: dudong@tsinghua.edu.cn [Department of Mechanical Engineering, Tsinghua University, Beijing 100084 (China); Key Laboratory for Advanced Materials Processing Technology, Ministry of Education P. R. China, Beijing 100084 (China); Huang, An; Chang, Baohua; Han, Zandong [Department of Mechanical Engineering, Tsinghua University, Beijing 100084 (China); Key Laboratory for Advanced Materials Processing Technology, Ministry of Education P. R. China, Beijing 100084 (China); He, Ayada [Shanghai Electric Power Generation Group Shanghai Generator Works, Shanghai 200240 (China)

    2016-08-15

    Highlights: • A detection system for locations of pebbles transported in pipes is introduced. • The detection system is based on vibration signal processing, which is original. • The characteristics of the vibration signals of the pipe are analyzed. • The experiment shows that the detection results are accurate. • The research provides an important basis for the design of the reactor. - Abstract: Pebble-bed high temperature gas-cooled reactors have many advantages such as inherent safety, high efficiency, etc., and have been considered as a candidate for Generation IV nuclear reactors. During the operation of the reactor, there are thousands of fuel pebbles transported in the pipes outside the core by gravity and helium flow. The pattern of the pipes which consist of straight and arc sections is very complex. When a fuel pebble is transported, it will constantly collide with the pipes, especially in the arc sections. The collisions will lead to the vibration of the pipes. This paper aims to provide a detection system for the location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing. Before the reactor is running, the system acquires the vibration signals of several key sections by sensors. Then the frequency characteristics of the signals are obtained by joint time–frequency analysis. When the reactor is running, the system detects the signals and processes them based on their frequency characteristics in real time. According to the results of the processing, the system can correctly judge whether the fuel pebble has passed through the section and records the time of the passing. The experiment validates the accuracy and reliability of the detection results. In this way, the operational condition of the reactor can be monitored so that the normal running of the reactor can be ensured. Additionally, the detection data are of great significance to the evaluation and optimization of the reactor performance

  17. Detection system for location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing

    International Nuclear Information System (INIS)

    Liu, Hongbing; Du, Dong; Huang, An; Chang, Baohua; Han, Zandong; He, Ayada

    2016-01-01

    Highlights: • A detection system for locations of pebbles transported in pipes is introduced. • The detection system is based on vibration signal processing, which is original. • The characteristics of the vibration signals of the pipe are analyzed. • The experiment shows that the detection results are accurate. • The research provides an important basis for the design of the reactor. - Abstract: Pebble-bed high temperature gas-cooled reactors have many advantages such as inherent safety, high efficiency, etc., and have been considered as a candidate for Generation IV nuclear reactors. During the operation of the reactor, there are thousands of fuel pebbles transported in the pipes outside the core by gravity and helium flow. The pattern of the pipes which consist of straight and arc sections is very complex. When a fuel pebble is transported, it will constantly collide with the pipes, especially in the arc sections. The collisions will lead to the vibration of the pipes. This paper aims to provide a detection system for the location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing. Before the reactor is running, the system acquires the vibration signals of several key sections by sensors. Then the frequency characteristics of the signals are obtained by joint time–frequency analysis. When the reactor is running, the system detects the signals and processes them based on their frequency characteristics in real time. According to the results of the processing, the system can correctly judge whether the fuel pebble has passed through the section and records the time of the passing. The experiment validates the accuracy and reliability of the detection results. In this way, the operational condition of the reactor can be monitored so that the normal running of the reactor can be ensured. Additionally, the detection data are of great significance to the evaluation and optimization of the reactor performance

  18. Dynamic analysis and application of fuel elements pneumatic transportation in a pebble bed reactor

    International Nuclear Information System (INIS)

    Liu, Hongbing; Du, Dong; Han, Zandong; Zou, Yirong; Pan, Jiluan

    2015-01-01

    Almost 10,000 spherical fuel elements are transported pneumatically one by one in the pipeline outside the core of a pebble bed reactor every day. Any failure in the transportation will lead to the shutdown of the reactor, even safety accidents. In order to ensure a stable and reliable transportation, it's of great importance to analyze the motion and force condition of the fuel element. In this paper, we focus on the dynamic analysis of the pneumatic transportation of the fuel element and derive kinetic equations. Then we introduce the design of the transportation pipeline. On this basis we calculate some important data such as the velocity of the fuel element, the force between the fuel element and the pipeline and the efficiency of the pneumatic transportation. Then we analyze these results and provide some suggestions for the design of the pipeline. The experiment was carried out on an experimental platform. The velocities of the fuel elements were measured. The experimental results were consistent with and validated the theoretical analysis. The research may offer the basis for the design of the transportation pipeline and the optimization of the fuel elements transportation in a pebble bed reactor. - Highlights: • The kinetic equations of the fuel element in pneumatic transportation are derived. • The dynamic characteristics of the fuel element are analyzed. • Some important parameters are calculated based on the kinetic equations. • The experimental results were consistent with the analysis and verified the analysis. • This paper may offer an important guide to the research of a pebble bed reactor

  19. The effects of temperatures on the pebble flow in a pebble bed high temperature reactor

    International Nuclear Information System (INIS)

    Sen, R. S.; Cogliati, J. J.; Gougar, H. D.

    2012-01-01

    The core of a pebble bed high temperature reactor (PBHTR) moves during operation, a feature which leads to better fuel economy (online refueling with no burnable poisons) and lower fuel stress. The pebbles are loaded at the top and trickle to the bottom of the core after which the burnup of each is measured. The pebbles that are not fully burned are recirculated through the core until the target burnup is achieved. The flow pattern of the pebbles through the core is of importance for core simulations because it couples the burnup distribution to the core temperature and power profiles, especially in cores with two or more radial burnup 'zones '. The pebble velocity profile is a strong function of the core geometry and the friction between the pebbles and the surrounding structures (other pebbles or graphite reflector blocks). The friction coefficient for graphite in a helium environment is inversely related to the temperature. The Thorium High Temperature Reactor (THTR) operated in Germany between 1983 and 1989. It featured a two-zone core, an inner core (IC) and outer core (OC), with different fuel mixtures loaded in each zone. The rate at which the IC was refueled relative to the OC in THTR was designed to be 0.56. During its operation, however, this ratio was measured to be 0.76, suggesting the pebbles in the inner core traveled faster than expected. It has been postulated that the positive feedback effect between inner core temperature, burnup, and pebble flow was underestimated in THTR. Because of the power shape, the center of the core in a typical cylindrical PBHTR operates at a higher temperature than the region next to the side reflector. The friction between pebbles in the IC is lower than that in the OC, perhaps causing a higher relative flow rate and lower average burnup, which in turn yield a higher local power density. Furthermore, the pebbles in the center region have higher velocities than the pebbles next to the side reflector due to the

  20. The effects of temperatures on the pebble flow in a pebble bed high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sen, R. S.; Cogliati, J. J.; Gougar, H. D. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2012-07-01

    The core of a pebble bed high temperature reactor (PBHTR) moves during operation, a feature which leads to better fuel economy (online refueling with no burnable poisons) and lower fuel stress. The pebbles are loaded at the top and trickle to the bottom of the core after which the burnup of each is measured. The pebbles that are not fully burned are recirculated through the core until the target burnup is achieved. The flow pattern of the pebbles through the core is of importance for core simulations because it couples the burnup distribution to the core temperature and power profiles, especially in cores with two or more radial burnup 'zones '. The pebble velocity profile is a strong function of the core geometry and the friction between the pebbles and the surrounding structures (other pebbles or graphite reflector blocks). The friction coefficient for graphite in a helium environment is inversely related to the temperature. The Thorium High Temperature Reactor (THTR) operated in Germany between 1983 and 1989. It featured a two-zone core, an inner core (IC) and outer core (OC), with different fuel mixtures loaded in each zone. The rate at which the IC was refueled relative to the OC in THTR was designed to be 0.56. During its operation, however, this ratio was measured to be 0.76, suggesting the pebbles in the inner core traveled faster than expected. It has been postulated that the positive feedback effect between inner core temperature, burnup, and pebble flow was underestimated in THTR. Because of the power shape, the center of the core in a typical cylindrical PBHTR operates at a higher temperature than the region next to the side reflector. The friction between pebbles in the IC is lower than that in the OC, perhaps causing a higher relative flow rate and lower average burnup, which in turn yield a higher local power density. Furthermore, the pebbles in the center region have higher velocities than the pebbles next to the side reflector due to the

  1. Advanced Core Design And Fuel Management For Pebble-Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2004-10-01

    A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

  2. Feasibility of Thorium Fuel Cycles in a Very High Temperature Pebble-Bed Hybrid System

    Directory of Open Access Journals (Sweden)

    L.P. Rodriguez

    2015-08-01

    Full Text Available Nuclear energy presents key challenges to be successful as a sustainable energy source. Currently, the viability of the use thorium-based fuel cycles in an innovative nuclear energy generation system is being investigated in order to solve these key challenges. In this work, the feasibility of three thorium-based fuel cycles (232Th-233U, 232Th-239Pu, and 232Th-U in a hybrid system formed by a Very High Temperature Pebble-Bed Reactor (VHTR and two Pebble-Bed Accelerator Driven Systems (ADSs was evaluated using parameters related to the neutronic behavior such as nuclear fuel breeding, minor actinide stockpile, the energetic contribution of each fissile isotope, and the radiotoxicity of the long lived wastes. These parameters were used to compare the fuel cycles using the well-known MCNPX ver. 2.6e computational code. The results obtained confirm that the 232Th-233U fuel cycle is the best cycle for minimizing the production of plutonium isotopes and minor actinides. Moreover, the inclusion of the second stage in the ADSs demonstrated the possibility of extending the burnup cycle duration and reducing the radiotoxicity of the discharged fuel from the VHTR.

  3. Development and engineering plan for graphite spent fuels conditioning program

    International Nuclear Information System (INIS)

    Bendixsen, C.L.; Fillmore, D.L.; Kirkham, R.J.; Lord, D.L.; Phillips, M.B.; Pinto, A.P.; Staiger, M.D.

    1993-09-01

    Irradiated (or spent) graphite fuel stored at the Idaho Chemical Processing Plant (ICPP) includes Fort St. Vrain (FSV) reactor and Peach Bottom reactor spent fuels. Conditioning and disposal of spent graphite fuels presently includes three broad alternatives: (1) direct disposal with minimum fuel packaging or conditioning, (2) mechanical disassembly of spent fuel into high-level waste and low-level waste portions to minimize geologic repository requirements, and (3) waste-volume reduction via burning of bulk graphite and other spent fuel chemical processing of the spent fuel. A multi-year program for the engineering development and demonstration of conditioning processes is described. Program costs, schedules, and facility requirements are estimated

  4. Method of manufacturing a graphite coated fuel can

    International Nuclear Information System (INIS)

    Saito, Koichi; Uchida, Shunsuke.

    1984-01-01

    Purpose: To improve the close bondability and homogeneity of a graphite coating formed at the inner surface of a fuel can. Method: A coating containing graphite dispersed in a volatile organic solvent is used and a graphite coating is formed to the inner surface of a fuel can by way of a plunger method. After applying graphite coating, an inert gas is caused to flow at a certain flow rate to the inside of the fuel can horizontally rotaged so that gassification and evaporation of the volatile organic solvent contained in the graphite coating may be promoted. Since drying of the graphite coating coated to the inner surface of the fuel can thus be controlled, a graphite coating with satisfactory close bondability and homogeneity can be formed. (Kawakami, Y.)

  5. Gas Turbine High Temperature Gas (Helium) Reactor Using Pebble Bed Fuel Derived from Spent Fuel

    International Nuclear Information System (INIS)

    Cole, Quentin

    2013-01-01

    Project goals: Build on the $1B investment spent during the NGNP Project for the only true Inherently Safe Small Modular Reactor Design – the only SMR design that can make this claim due to negative temperature coefficient of reactivity - no containment required – less construction cost. NPMC in Partnership with Pebble Bed Modular Group, a fully owned subsidiary of Eskom, RSA to Factory Build Complete Plant in Modular Sections at Factory Site in Oswego, NY for transport to site by rail or shipping for world wide export. NPMC will provide Project and Construction Management of all new builds from plant sites through construction, commissioning and startup using local labor. License and Construct ion of spent fuel processing facility in both NY and South Africa using Proven Technology. Ultimate goals of project: 1. Award of the 2013 US DOE Innovative SMR $452M cost share grant for US NRC License Certification 2.Build Full Scale Demonstration Plant at Koeburg, RSA with World Bank Funding managed by NPMC in collaboration with our legal firm, Haynes and Boone LLP 3. Take Plant Orders Immediately (10% Down Payment) 4. Form Strategic Alliance with Domestic and/or International Utility

  6. Features of spherical uranium-graphite HTGR fuel elements control

    International Nuclear Information System (INIS)

    Kreindlin, I.I.; Oleynikov, P.P.; Shtan, A.S.

    1985-01-01

    Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described

  7. Features of spherical uranium-graphite HTGR fuel elements control

    Energy Technology Data Exchange (ETDEWEB)

    Kreindlin, I I; Oleynikov, P P; Shtan, A S

    1985-07-01

    Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described.

  8. Graphite

    Science.gov (United States)

    Robinson, Gilpin R.; Hammarstrom, Jane M.; Olson, Donald W.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

    2017-12-19

    Graphite is a form of pure carbon that normally occurs as black crystal flakes and masses. It has important properties, such as chemical inertness, thermal stability, high electrical conductivity, and lubricity (slipperiness) that make it suitable for many industrial applications, including electronics, lubricants, metallurgy, and steelmaking. For some of these uses, no suitable substitutes are available. Steelmaking and refractory applications in metallurgy use the largest amount of produced graphite; however, emerging technology uses in large-scale fuel cell, battery, and lightweight high-strength composite applications could substantially increase world demand for graphite.Graphite ores are classified as “amorphous” (microcrystalline), and “crystalline” (“flake” or “lump or chip”) based on the ore’s crystallinity, grain-size, and morphology. All graphite deposits mined today formed from metamorphism of carbonaceous sedimentary rocks, and the ore type is determined by the geologic setting. Thermally metamorphosed coal is the usual source of amorphous graphite. Disseminated crystalline flake graphite is mined from carbonaceous metamorphic rocks, and lump or chip graphite is mined from veins in high-grade metamorphic regions. Because graphite is chemically inert and nontoxic, the main environmental concerns associated with graphite mining are inhalation of fine-grained dusts, including silicate and sulfide mineral particles, and hydrocarbon vapors produced during the mining and processing of ore. Synthetic graphite is manufactured from hydrocarbon sources using high-temperature heat treatment, and it is more expensive to produce than natural graphite.Production of natural graphite is dominated by China, India, and Brazil, which export graphite worldwide. China provides approximately 67 percent of worldwide output of natural graphite, and, as the dominant exporter, has the ability to set world prices. China has significant graphite reserves, and

  9. Preliminary analysis on in-core fuel management optimization of molten salt pebble-bed reactor

    International Nuclear Information System (INIS)

    Xia Bing; Jing Xingqing; Xu Xiaolin; Lv Yingzhong

    2013-01-01

    The Nuclear Hot Spring (NHS) is a molten salt pebble-bed reactor featured by full power natural circulation. The unique horizontal coolant flow of the NHS demands the fuel recycling schemes based on radial zoning refueling and the corresponding method of fuel management optimization. The local searching algorithm (LSA) and the simulated annealing algorithm (SAA), the stochastic optimization methods widely used in the refueling optimization problems in LWRs, were applied to the analysis of refueling optimization of the NHS. The analysis results indicate that, compared with the LSA, the SAA can survive the traps of local optimized solutions and reach the global optimized solution, and the quality of optimization of the SAA is independent of the choice of the initial solution. The optimization result gives excellent effects on the in-core power flattening and the suppression of fuel center temperature. For the one-dimensional zoning refueling schemes of the NHS, the SAA is an appropriate optimization method. (authors)

  10. 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

  11. TAPIR, Thermal Analysis of HTGR with Graphite Sleeve Fuel Elements

    International Nuclear Information System (INIS)

    Weicht, U.; Mueller, W.

    1983-01-01

    1 - Nature of the physical problem solved: Thermal analysis of a reactor core containing internally and/or externally gas cooled prismatic fuel elements of various geometries, rating, power distribution, and material properties. 2 - Method of solution: A fuel element in this programme is regarded as a sector of a fuelled annulus with graphite sleeves of any shape on either side and optional annular gaps between fuel and graphite and/or within the graphite. It may have any centre angle and the fuelled annulus may become a solid cylindrical rod. Heat generation in the fuel is assumed to be uniform over the cross section and peripheral heat flux into adjacent sectors is ignored. Fuel elements and coolant channels are treated separately, then linked together to fit a specified pattern. 3 - Restrictions on the complexity of the problem: Maxima of: 50 fuel elements; 50 cooled channels; 25 fuel geometries; 25 coolant channel geometries; 10 axial power distributions; 10 graphite conductivities

  12. (Fuel, fission product, and graphite technology)

    Energy Technology Data Exchange (ETDEWEB)

    Stansfield, O.M.

    1990-07-25

    Travel to the Forschungszentrum (KFA) -- Juelich described in this report was for the purpose of participating in the annual meeting of subprogram managers for the US/DOE Umbrella Agreement for Fuel, Fission Product, and Graphite Technology. At this meeting the highlights of the cooperative exchange were reviewed for the time period June 1989 through June 1990. The program continues to contribute technology in an effective way for both countries. Revision 15 of the Subprogram Plan will be issued as a result of the meeting. There was interest expressed by KFA management in the level of support received from the NPR program and in potential participation in the COMEDIE loop experiment being conducted at the CEA.

  13. KüFA safety testing of HTR fuel pebbles irradiated in the High Flux Reactor in Petten

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, O., E-mail: oliver.seeger@rwth-aachen.de [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Laurie, M., E-mail: mathias.laurie@ec.europa.eu [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Abjani, A. El; Ejton, J.; Boudaud, D.; Freis, D.; Carbol, P.; Rondinella, V.V. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Fütterer, M. [European Commission, Joint Research Centre (JRC), Institute for Energy and Transport (IET), Nuclear Reactor Integrity Assessment and Knowledge Management Unit, PO Box 2, 1755 ZG Petten (Netherlands); Allelein, H.-J. [Lehrstuhl für Reaktorsicherheit und -technik an der RWTH Aachen, Kackertstraße 9, 52072 Aachen (Germany)

    2016-09-15

    The Cold Finger Apparatus (KühlFinger-Apparatur—KüFA) in operation at JRC-ITU is designed to experimentally scrutinize the effects of Depressurization LOss of Forced Circulation (D-LOFC) accident scenarios on irradiated High Temperature Reactor (HTR) fuel pebbles. Up to 1600 °C, the reference maximum temperature for these accidents, high-quality German HTR fuel pebbles have already demonstrated a small fission product release. This paper discusses and compares the releases obtained from KüFA-testing the pebbles HFR-K5/3 and HFR-EU1/3, which were both irradiated in the High Flux Reactor (HFR) in Petten. We present the time-dependent fractional release of the volatile fission product {sup 137}Cs as well as the fission gas {sup 85}Kr for both pebbles. For HFR-EU1/3 the isotopes {sup 134}Cs and {sup 154}Eu as well as the shorter-lived {sup 110m}Ag have also been measured. A detailed description of the experimental setup and its accuracy is given. The data for the recently tested pebbles is discussed in the context of previous results.

  14. Development of Monte Carlo-based pebble bed reactor fuel management code

    International Nuclear Information System (INIS)

    Setiadipura, Topan; Obara, Toru

    2014-01-01

    Highlights: • A new Monte Carlo-based fuel management code for OTTO cycle pebble bed reactor was developed. • The double-heterogeneity was modeled using statistical method in MVP-BURN code. • The code can perform analysis of equilibrium and non-equilibrium phase. • Code-to-code comparisons for Once-Through-Then-Out case were investigated. • Ability of the code to accommodate the void cavity was confirmed. - Abstract: A fuel management code for pebble bed reactors (PBRs) based on the Monte Carlo method has been developed in this study. The code, named Monte Carlo burnup analysis code for PBR (MCPBR), enables a simulation of the Once-Through-Then-Out (OTTO) cycle of a PBR from the running-in phase to the equilibrium condition. In MCPBR, a burnup calculation based on a continuous-energy Monte Carlo code, MVP-BURN, is coupled with an additional utility code to be able to simulate the OTTO cycle of PBR. MCPBR has several advantages in modeling PBRs, namely its Monte Carlo neutron transport modeling, its capability of explicitly modeling the double heterogeneity of the PBR core, and its ability to model different axial fuel speeds in the PBR core. Analysis at the equilibrium condition of the simplified PBR was used as the validation test of MCPBR. The calculation results of the code were compared with the results of diffusion-based fuel management PBR codes, namely the VSOP and PEBBED codes. Using JENDL-4.0 nuclide library, MCPBR gave a 4.15% and 3.32% lower k eff value compared to VSOP and PEBBED, respectively. While using JENDL-3.3, MCPBR gave a 2.22% and 3.11% higher k eff value compared to VSOP and PEBBED, respectively. The ability of MCPBR to analyze neutron transport in the top void of the PBR core and its effects was also confirmed

  15. Renewable side reflector structure for a pebble bed high temperature reactor

    International Nuclear Information System (INIS)

    Martin, Roger.

    1977-01-01

    The description is given of a renewable side reflector structure for a pebble bed high temperature reactor of the kind comprising a cylindrical graphite vessel constituting the neutron reflector, this vessel being filled with graphite pebbles containing the nuclear fuel and enclosed in a concrete protective containment. The internal peripheral area of the vessel is constituted by a line of adjacent graphite rods mounted so that they can rotate about their longitudinal axis and manoeuvrable from outside the concrete containment by means of a shaft passing into it [fr

  16. Milling Behavior of Matrix Graphite Powders with Different Binder Materials in HTGR Fuel Element Fabrication: I. Variation in Particle Size Distribution

    International Nuclear Information System (INIS)

    Lee, Young Woo; Cho, Moon Sung

    2011-01-01

    The fuel element for HTGR is manufactured by mixing coated fuel particles with matrix graphite powder and forming into either pebble type or cylindrical type compacts depending on their use in different HTGR cores. The coated fuel particle, the so-called TRISO particle, consists of 500-μm spherical UO 2 particles coated with the low density buffer Pyrolytic Carbon (PyC) layer, the inner and outer high density PyC layer and SiC layer sandwiched between the two inner and outer PyC layers. The coated TRISO particles are mixed with a matrix graphite powder properly prepared and pressed into a spherical shape or a cylindrical compact finally heat-treated at about 1900 .deg. C. These fuel elements can have different sizes and forms of compact. The basic steps for manufacturing a fuel element include preparation of graphite matrix powder, overcoating the fuel particles, mixing the fuel particles with a matrix powder, carbonizing green compact, and the final high-temperature heat treatment of the carbonized fuel compact. In order to develop a fuel compact fabrication technology, it is important to develop a technology to prepare the matrix graphite powder (MGP) with proper characteristics, which has a strong influence on further steps and the material properties of fuel element. In this work, the milling behavior of matrix graphite powder mixture with different binder materials and their contents was investigated by analyzing the change in particle size distribution with different milling time

  17. Low temperature chemical processing of graphite-clad nuclear fuels

    Science.gov (United States)

    Pierce, Robert A.

    2017-10-17

    A reduced-temperature method for treatment of a fuel element is described. The method includes molten salt treatment of a fuel element with a nitrate salt. The nitrate salt can oxidize the outer graphite matrix of a fuel element. The method can also include reduced temperature degradation of the carbide layer of a fuel element and low temperature solubilization of the fuel in a kernel of a fuel element.

  18. Engineering solutions for a reflector change concept in the high-temperature reactor with pebble bed core and OTTO-fueling

    International Nuclear Information System (INIS)

    Kasper, K.J.

    1975-06-01

    In the field of reactor engineering an increasing tendency is visible towards a 'repairable reactor'. In the construction of the HTR with spherical fuel elements this fact should already be taken into account at an early stage. Additionally it is possible that in connection with the OTTO-fueling load conditions for the graphite reflector could result which are locally not far away from limiting values. Therefore the removability of the reflector is included in the reactor construction as an accompanying technical step of the physical lay-out of the core. The core arrangements, realized for HTR until recently, are discussed as well as the properties of the graphites used and the operating conditions in the reactors are stated. At the example of the PR 3,000 proposals are offered for the construction of a removable side and top reflector for a pebble bed reactor. Hereby a solution was found which, on one hand allows the changing of the reflector and on the other hand requires no significant increase of the costs for the reactor assembly. Moreover the requirements of reactor operation and of repairability are satisfied in an optimal manner. (orig.) [de

  19. Analysis of the impact of random summing on passive assay of pebble bed reactor fuel using gamma-ray spectrometry

    Science.gov (United States)

    Chen, J.; Hawari, A. I.

    2007-08-01

    Pebble bed reactors (PBR) are characterized by multi-pass fuel systems in which spherical fuel pebbles are circulated through the core until they reach a proposed burnup limit. The fuel is assayed on-line to ensure that the burnup limit is not breached. However, random summing effects can impact the response of the burnup measurement system and result in distortions that degrade the accuracy of the assay results. Monte Carlo analysis was performed to estimate the magnitude and effect of random summing on the absolute and relative indicators that have been identified as usable in on-line assay. For a throughput rate of 10 5 counts/s and trapezoidal pulse shaping of the signals, the results show that absolute indicators suffer from severe distortions due to this effect. Relative indicators are found to be resistant to random summing with the deviation in the ratio of peak areas remaining less than 5-15% depending on pulse width.

  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. HTR-PROTEUS pebble bed experimental program cores 9 & 10: columnar hexagonal point-on-point packing with a 1:1 moderator-to-fuel pebble ratio

    Energy Technology Data Exchange (ETDEWEB)

    Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-03-01

    PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.

  3. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 9 & 10: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2013-03-01

    PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.

  4. US/FRG joint report on the pebble bed high temperature reactor resource conservation potential and associated fuel cycle costs

    International Nuclear Information System (INIS)

    Teuchert, E.; Ruetten, H.J.; Worley, B.A.; Vondy, D.R.

    1979-11-01

    Independent analyses at ORNL and KFA have led to the general conclusion that the flexibility in design and operation of a high-temperature gas-cooled pebble-bed reactor (PBR) can result in favorable ore utilization and fuel costs in comparison with other reactor types, in particular, with light-water reactors (LWRs). Fuel reprocessign and recycle show considerable promise for reducing ore consumption, and even the PBR throwaway cycle is competitive with fuel recycle in an LWR. The best performance results from the use of highly enriched fuel. Proliferation-resistant measures can be taken using medium-enriched fuel at a modest ore penalty, while use of low-enriched fuel would incur further ore penalty. Breeding is possible but net generation of fuel at a significant rate would be expensive, becoming more feasible as ore costs increase substantially. The 233 U inventory for a breeder could be produced by prebreeders using 235 U fuel

  5. Corrosion of graphite composites in phosphoric acid fuel cells

    Science.gov (United States)

    Christner, L. G.; Dhar, H. P.; Farooque, M.; Kush, A. K.

    1986-01-01

    Polymers, polymer-graphite composites and different carbon materials are being considered for many of the fuel cell stack components. Exposure to concentrated phosphoric acid in the fuel cell environment and to high anodic potential results in corrosion. Relative corrosion rates of these materials, failure modes, plausible mechanisms of corrosion and methods for improvement of these materials are investigated.

  6. Graphite behaviour in relation to the fuel element design

    Energy Technology Data Exchange (ETDEWEB)

    Everett, M. R. [OECD High Temperature Reactor Project Dragon, Winfrith (United Kingdom); Manzel, R. [OECD High Temperature Reactor Project Dragon, Winfrith (United Kingdom); Blackstone, R. [Reactor Centrum, Petten (Netherlands); Delle, W. [Kernforschungsanlage, Juelich (Germany); Lungagnani, V. [Joint Nuclear Research Centre, Euratom, Petten (Netherlands); Krefeld, R. [Joint Nuclear Research Centre, Euratom, Petten (Netherlands)

    1969-09-01

    The first designs of H.T.R. power reactors will probably use a Gilsocarbon based graphite for both the moderator/carrier blocks and for the fuel tubes. The initial physical properties and changes of dimensions, thermal expansion coefficient, Young*s modulus, and thermal conductivity on irradiation of Gilsocarbon graphites to typical reactor dwell-time fast neutron doses of 4 * 1021 cm -2 Ni dose Dido equivalent are given and values for the irradiation creep constant are presented. The influence of these property changes and those of chemical corrosion are considered briefly in relation to the present fuel element designs. The selection of an eventual less costly replacement graphite for Gilsocarbon graphite is discussed in terms of materials properties.

  7. Structure and influence factors of fuel cycle costs of pebble bed HTRs with OTTO-fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Jacke, S.

    1975-06-15

    The study in this paper can be divided into two parts. The first part deals with the analysis of the structure of the fuel cycle costs of today in 1974. A comparison is made between two pebble bed HTRs with OTTO-refueling-management (once-through) and a LWR of the type Biblis A. The two HTRs use different fuels: The one low-enriched Uranium (LOTTO), the other high-enriched Uranium and Thorium (TOTTO). The analysis of the structure of the fuel cycle costs consists of a discussion of the most important input parameters, and a comparison of each cost item. This study was made without adjustment of the core design to the changing market conditions. It is quite natural that an adaptation of the moderation ratio, of the conversion ratio, of the enrichment level, and of the burn-up may lower the fuel cycle costs. But the differences cannot be very important, and the results of this examination may remain valid, even on best adjustment conditions.

  8. Researchers solve big mysteries of pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shams, Afaque; Roelofs, Ferry; Komen, E.M.J. [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Baglietto, Emilio [Massachusetts Institute of Technology, Cambridge, MA (United States). Dept. of Nuclear Science and Engineering; Sgro, Titus [CD-adapco, London (United Kingdom). Technical Marketing

    2014-03-15

    The PBR is one type of High Temperature Reactors, which allows high temperature work while preventing the fuel from melting (bringing huge safety margins to the reactor) and high electricity efficiency. The design is also highly scalable; a plant could be designed to be as large or small as needed, and can even be made mobile, allowing it to be used onboard a ship. In a PBR, small particles of nuclear fuel, embedded in a moderating graphite pebble, are dropped into the reactor as needed. At the bottom, the pebbles can be removed simply by opening a small hatch and letting gravity pull them down. To cool the reactor and create electricity, helium gas is pumped through the reactor to pull heat out which is then run through generators. One of the most difficult problems to deal with has been the possible appearance of local temperature hotspots within the pebble bed heating to the point of melting the graphite moderators surrounding the fuel. Obviously, constructing a reactor and experimenting to investigate this possibility is out of the question. Instead, nuclear engineers have been attempting to simulate a PBR with various CFD codes. The thermo-dynamic analysis to simulate realistic conditions in a pebble bed are described and the results are shown. (orig.)

  9. Preliminary design study of pebble bed reactor HTR-PM base using once-through-then-out fuel recirculation

    International Nuclear Information System (INIS)

    Topan Setiadipura; Jupiter S Pane; Zuhair

    2016-01-01

    Pebble Bed Reactor (PBR) is one of the advanced reactor type implementing strong passive safety feature. In this type of design has the potential to do a cogeneration useful for the treatment of various minerals in various islands in Indonesia. The operation of the PBR can be simplified by implementing once-through-then-out (OTTO) fuel recirculation scheme in which pebble fuel only pass the core once time. The purpose of this research is to understand quantitative influence of the changing of fuel element recirculation on the PBR core performance and to find preliminary optimization design of PBR type reactor with OTTO recirculation scheme. PEBBED software was used to find PBR equilibrium core. The calculation result gives quantitative data on the impact of implementing a different fuel recirculation, especially using OTTO scheme. Furthermore, an early optimized PBR design based on HTR-PM using OTTO scheme was obtained where the power must be downgraded into 115 MWt in order to preserve the safety feature. The simplicity of the reactor operation and the reduction of reactor component with OTTO scheme still make this early optimized design an interesting alternative design, despite its power reduction from the reference design. (author)

  10. Characterization of graphite-matrix pulsed reactor fuels

    International Nuclear Information System (INIS)

    Karnes, C.H.; Marion, R.H.

    1976-01-01

    The performance of the Annular Core Pulsed Reactor (ACPR) is being upgraded in order to accommodate higher fluence experiments for fast reactor fuel element transient and safety studies. The increased fluence requires a two-zone core with the inner zone containing fuel having a high enthalpy and the capability of withstanding very high temperatures during both pulsed and steady state operation. Because the fuel is subjected to a temperature risetime of 2 to 5 ms and to a large temperature difference across the diameter, fracture due to thermal stresses is the primary failure mode. One of the fuels considered for the high enthalpy inner region is a graphite-matrix fuel containing a dispersion of uranium--zirconium carbide solid solution particles. A program was initiated to optimize the development of this class of fuel. This summary presents results on formulations of fuel which have been fabricated by the Materials Technology Group of the Los Alamos Scientific Laboratory

  11. Safeguards-by-Design: Guidance for High Temperature Gas Reactors (HTGRs) With Pebble Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Philip Casey Durst; Mark Schanfein

    2012-08-01

    The following is a guidance document from a series prepared for the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA), under the Next Generation Safeguards Initiative (NGSI), to assist facility designers and operators in implementing international Safeguards-by-Design (SBD). SBD has two main objectives: (1) to avoid costly and time consuming redesign work or retrofits of new nuclear fuel cycle facilities and (2) to make the implementation of international safeguards more effective and efficient at such facilities. In the long term, the attainment of these goals would save industry and the International Atomic Energy Agency (IAEA) time, money, and resources and be mutually beneficial. This particular safeguards guidance document focuses on pebble fuel high temperature gas reactors (HTGR). The purpose of the IAEA safeguards system is to provide credible assurance to the international community that nuclear material and other specified items are not diverted from peaceful nuclear uses. The safeguards system consists of the IAEA’s statutory authority to establish safeguards; safeguards rights and obligations in safeguards agreements and additional protocols; and technical measures implemented pursuant to those agreements. Of foremost importance is the international safeguards agreement between the country and the IAEA, concluded pursuant to the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). According to a 1992 IAEA Board of Governors decision, countries must: notify the IAEA of a decision to construct a new nuclear facility as soon as such decision is taken; provide design information on such facilities as the designs develop; and provide detailed design information based on construction plans at least 180 days prior to the start of construction, and on "as-built" designs at least 180 days before the first receipt of nuclear material. Ultimately, the design information will be captured in an IAEA Design Information

  12. Numerical Simulation of a Coolant Flow and Heat Transfer in a Pebble Bed Reactor

    International Nuclear Information System (INIS)

    In, Wang-Kee; Kim, Min-Hwan; Lee, Won-Jae

    2008-01-01

    Pebble Bed Reactor(PBR) is one of the very high temperature gas cooled reactors(VHTR) which have been reviewed in the Generation IV International Forum as potential sources for future energy needs, particularly for a hydrogen production. The pebble bed modular reactor(PBMR) exhibits inherent safety features due to the low power density and the large amount of graphite present in the core. PBR uses coated fuel particles(TRISO) embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the PBR core during a reactor operation and the coolant flows around randomly distributed spheres. For the reliable operation and the safety of the PBR, it is important to understand the coolant flow structure and the fuel pebble temperature in the PBR core. There have been few experimental and numerical studies to investigate the fluid and heat transfer phenomena in the PBR core. The objective of this paper is to predict the fluid and heat transfer in the PBR core. The computational fluid dynamics (CFD) code, STAR-CCM+(V2.08) is used to perform the CFD analysis using the design data for the PBMR400

  13. Progress on the development of a new fuel management code to simulate the movement of pebble and block type fuel elements in a very high temperature reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Xhonneux, Andre, E-mail: a.xhonneux@fz-juelich.de [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology, RWTH-Aachen, 52064 Aachen (Germany); Kasselmann, Stefan; Rütten, Hans-Jochem [Forschungszentrum Jülich, 52425 Jülich (Germany); Becker, Kai [Institute for Reactor Safety and Reactor Technology, RWTH-Aachen, 52064 Aachen (Germany); Allelein, Hans-Josef [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology, RWTH-Aachen, 52064 Aachen (Germany)

    2014-05-01

    The history of gas-cooled high-temperature reactor prototypes in Germany is closely related to Forschungszentrum Jülich and its “Institute of Nuclear Waste Disposal and Reactor Safety (IEK-6)”. A variety of computer codes have been developed, validated and optimized to simulate the different safety and operational aspects of V/HTR. In order to overcome the present limitations of these codes and to exploit the advantages of modern computer clusters, a project has been initiated to integrate these individual programs into a consistent V/HTR code package (VHCP) applying state-of-the-art programming techniques and standards. One important aspect in the simulation of a V/HTR is the modeling of a continuous moving pebble bed or the periodic rearrangement of prismatic block type fuel. Present models are either too coarse to take special issues (e.g. pebble piles) into account or are too detailed and therefore too time consuming to be applicable in the HCP. The new Software for Handling Universal Fuel Elements (SHUFLE) recently being developed is well suited to close this gap. Although at first the code has been designed for pebble bed reactors, it can in principal be applied to all other types of nuclear fuel. The granularity of the mesh grid meets the requirements to consider these special issues while keeping the used computing power within reasonable limits. New features are for example the possibility to consider azimuthally differing flow velocities in the case of a pebble bed reactor or individual void factors to simulate effects to seismic events. The general idea behind this new approach to the simulation of pebble bed reactors is the following: In the preprocessing step, experimental flow lines or flow lines simulated by more detailed codes serve as an input. For each radial mesh column a representative flow line is then determined by interpolation. These representative flow lines are finally mapped to a user defined rectangular grid forming chains of meshes

  14. Progress on the development of a new fuel management code to simulate the movement of pebble and block type fuel elements in a very high temperature reactor core

    International Nuclear Information System (INIS)

    Xhonneux, Andre; Kasselmann, Stefan; Rütten, Hans-Jochem; Becker, Kai; Allelein, Hans-Josef

    2014-01-01

    The history of gas-cooled high-temperature reactor prototypes in Germany is closely related to Forschungszentrum Jülich and its “Institute of Nuclear Waste Disposal and Reactor Safety (IEK-6)”. A variety of computer codes have been developed, validated and optimized to simulate the different safety and operational aspects of V/HTR. In order to overcome the present limitations of these codes and to exploit the advantages of modern computer clusters, a project has been initiated to integrate these individual programs into a consistent V/HTR code package (VHCP) applying state-of-the-art programming techniques and standards. One important aspect in the simulation of a V/HTR is the modeling of a continuous moving pebble bed or the periodic rearrangement of prismatic block type fuel. Present models are either too coarse to take special issues (e.g. pebble piles) into account or are too detailed and therefore too time consuming to be applicable in the HCP. The new Software for Handling Universal Fuel Elements (SHUFLE) recently being developed is well suited to close this gap. Although at first the code has been designed for pebble bed reactors, it can in principal be applied to all other types of nuclear fuel. The granularity of the mesh grid meets the requirements to consider these special issues while keeping the used computing power within reasonable limits. New features are for example the possibility to consider azimuthally differing flow velocities in the case of a pebble bed reactor or individual void factors to simulate effects to seismic events. The general idea behind this new approach to the simulation of pebble bed reactors is the following: In the preprocessing step, experimental flow lines or flow lines simulated by more detailed codes serve as an input. For each radial mesh column a representative flow line is then determined by interpolation. These representative flow lines are finally mapped to a user defined rectangular grid forming chains of meshes

  15. Optimized core design and fuel management of a pebble-bed type nuclear reactor

    NARCIS (Netherlands)

    Boer, B.

    2009-01-01

    The core design of a pebble-bed type Very High Temperature Reactor (VHTR) is optimized, aiming for an increase of the coolant outlet temperature to 1000 C, while retaining its inherent safety features. The VHTR has been selected by the international Generation IV research initiative as one of the

  16. KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis

    International Nuclear Information System (INIS)

    Shamasundar, B.I.; Fehrenbach, M.E.

    1981-05-01

    The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations

  17. Nuclear safeguards considerations for pebble bed reactors (PBRs)

    International Nuclear Information System (INIS)

    Moses, David L.

    2012-01-01

    Recent reports by the Department of Energy National Laboratories have discussed safeguards considerations for low enriched uranium (LEU)-fueled pebble bed reactors (PBRs) and the need for bulk accountancy of the plutonium in “used fuel.” These reports fail to account for the degree of plutonium dilution in the graphitized-carbon pebbles that is sufficient to meet the International Atomic Energy Agency (IAEA) “provisional” guidelines for termination of safeguards on “measured discards.” The thrust of this finding is not to terminate safeguards but to limit the need for specific accountancy of plutonium in stored used fuel. While the residual uranium in the used fuel is not sufficiently diluted to meet the IAEA provisional guidelines for termination of safeguards, the estimated quantities of the uranium minor isotopes 232 U and 236 U in the used fuel at the target burnup of ∼90 Gigawatt-days per metric ton (GWD/MT) exceed standard specification limits for reprocessed uranium and will require extensive blending with either natural uranium or uranium enrichment tails to dilute the 236 U content to fall within specification. Hence, the PBR used fuel is less desirable for commercial reprocessing and reuse than that from light water reactors. Also the PBR specific activity of a reprocessed uranium isotopic mixture and its A 2 values for effective dose limits if released in a dispersible form during a transportation accident are more limiting than the equivalent values for light-water-reactor used fuel at 55 GWD/MT without accounting for the presence of the principal carry-over fission product (technetium, 99 Tc) and plutonium contamination. Thus, the potentially recoverable uranium from PBR used fuel carries reactivity penalties and radiological penalties likely greater than those for reprocessed uranium from light water reactors. These factors impact the economics of reprocessing, but a more significant consideration is that reprocessing technologies for

  18. 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)

  19. Improved graphite matrix for coated-particle fuel

    International Nuclear Information System (INIS)

    Schell, D.H.; Davidson, K.V.

    1978-10-01

    An experimental process was developed to incorporate coated fuel particles in an extruded graphite matrix. This structure, containing 41 vol% particles, had a high matrix density, >1.6 g/cm 3 , and a matrix conductivity three to four times that of a pitch-injected fuel rod at 1775 K. Experiments were conducted to determine the uniformity of particle loadings in extrusions. Irradiation specimens were supplied for five tests in the High-Fluence Isotope Reactor at the Oak Ridge National Laboratory

  20. Study on the Calculation of Pebble-Bed Reactor Multiplication Factor As a Function of Fuel Kernel Radius at Various Enrichments

    International Nuclear Information System (INIS)

    Zuhair; Suwoto

    2009-01-01

    Main characteristics of PBR comes from utilization of coated particle fuels dispersed in pebble fuels . Because of vibration, fuel kernel can be grouped into cluster and in these cases, neutronic characteristics of pebble fuel significantly changes . In this study, cluster is modeled structural form consisting of uniform cubic cells with eight neighborhood TRISO particles . Neutronic characteristics was investigated by calculating pebble-bed reactor multiplication factor as a function of fuel kernel radius at various enrichments . The calculation results using MCNP5 code with ENDF/BVI neutron library show that k eff value depends on the average fuel radius and reaches its minimum when all kernels have the same radius, i.e. 0.0280 cm . With this radius, the total kernel surface area achieves maximum value . The dependence of k eff on fuel kernel radius decreases in relation to the increase in uranium enrichment . However, k eff value is not affected by fuel kernel radius when the uranium is 100% enriched . From these result, it can be concluded that, exception of uranium enrichment, the selection of fuel kernel radius should be considered thoroughly in designing a PBR, since this parameter provides significant influences on neutronic characteristics of the reactor. (author)

  1. Operating windows of pebble divertor

    International Nuclear Information System (INIS)

    Matsuhiro, K.; Isobe, M.; Ohtsuka, Y.; Ueda, Y.; Nishikawa, M.

    2001-01-01

    A marked feature of the pebble divertor is an effect by use of functional multi-layer coated pebble, which consists of a surface plasma facing layer, an intermediate tritium permeation barrier layer, and a kernel for heat removal. The dimensions, structure and the irradiation conditions of pebbles are the important issues for the development of the pebble divertor. From the view point of resistance of the induced thermal stress, the pebble is taken as small as possible in size. On the other hand, from the view point of the pumping performance, the suitable irradiation temperature range of the surface layer of pebble was estimated from the experiments and the numerical analysis. The pumping process enhanced by dynamic retention is available to extend the higher allowable irradiation temperature range from 900K to 1100K. As taking the temperature rise limitation due to pumping effect and the fractural strength due to the induced thermal stress limitation, it was found that the diameter of the pebble is possible to be 1-2 mm in about 20 MW/m 2 for the SiC kernel and 2-3 mm in less than 30 MW/m 2 for the graphite kernel. (author)

  2. Irradiation-induced dimensional changes of fuel compacts and graphite sleeves of OGL-1 fuel assemblies

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Minato, Kazuo; Kobayashi, Fumiaki; Tobita, Tsutomu; Kikuchi, Teruo; Kurobane, Shiro; Adachi, Mamoru; Fukuda, Kousaku

    1988-06-01

    Experimental data are summarized on irradiation-induced dimensional changes of fuel compacts and graphite sleeves of the first to ninth OGL-1 fuel assemblies. The range of fast-neutron fluence is up to 4 x 10 24 n/m 2 (E > 0.18 MeV); and that of irradiation temperature is 900 - 1400 deg C for fuel compacts and 800 - 1050 deg C for graphite sleeves. The dimensional change of the fuel compacts was shrinkage under these test conditions, and the shrinkage fraction increased almost linearly with fast-neutron fluence. The shrinkage fraction of the fuel compacts was larger by 20 % in the axial direction than in the radial direction. Influence of the irradiation temperature on the dimensional-change behavior of the fuel compacts was not observed clearly; presumably the influence was hidden by scatter of the data because of low level of the fast-neutron fluence and the resultant small dimensional changes. (author)

  3. Core homogenization method for pebble bed reactors

    International Nuclear Information System (INIS)

    Kulik, V.; Sanchez, R.

    2005-01-01

    This work presents a core homogenization scheme for treating a stochastic pebble bed loading in pebble bed reactors. The reactor core is decomposed into macro-domains that contain several pebble types characterized by different degrees of burnup. A stochastic description is introduced to account for pebble-to-pebble and pebble-to-helium interactions within a macro-domain as well as for interactions between macro-domains. Performance of the proposed method is tested for the PROTEUS and ASTRA critical reactor facilities. Numerical simulations accomplished with the APOLLO2 transport lattice code show good agreement with the experimental data for the PROTEUS reactor facility and with the TRIPOLI4 Monte Carlo simulations for the ASTRA reactor configuration. The difference between the proposed method and the traditional volume-averaged homogenization technique is negligible while only one type of fuel pebbles present in the system, but it grows rapidly with the level of pebble heterogeneity. (authors)

  4. Research on application of burnable poison in pebble bed HTR

    International Nuclear Information System (INIS)

    Wei Chunlin; Zhang Jian; Shan Wenzhi; Jing Xingqing

    2013-01-01

    Burnable poison in fuel ball was used in pebble bed high-temperature gas-cooled reactor (HTR) to optimize the shape and the peak factor of power distribution in certain conditions. Two options are available and evaluated, that is the homogeneous burnable poison in graphite matrix and burnable poison particles (BPPs) in fuel balls. Due to the absorption cross section of "1"0B, the depletion speed for homogeneous burnable poison is very fast, and difficult to control, on the other side, the depletion speed of BPPs can be optimized respecting to its size, and better shape and peak value of power distribution can be achieved. (authors)

  5. Surface-reconstructed graphite nanofibers as a support for cathode catalysts of fuel cells.

    Science.gov (United States)

    Gan, Lin; Du, Hongda; Li, Baohua; Kang, Feiyu

    2011-04-07

    Graphite nanofibers (GNFs), on which surface graphite edges were reconstructed into nano-loops, were explored as a cathode catalyst support for fuel cells. The high degree of graphitization, as well as the surface-reconstructed nano-loops that possess topological defects for uniform metal deposition, resulted in an improved performance of the GNF-supported Pt catalyst.

  6. Pebble-bed reactor

    International Nuclear Information System (INIS)

    Lohnert, G.; Mueller-Frank, U.; Heil, J.

    1976-01-01

    A pebble-bed nuclear reactor of large power rating comprises a container having a funnel-shaped bottom forming a pebble run-out having a centrally positioned outlet. A bed of downwardly-flowing substantially spherical nuclear fuel pebbles is positioned in the container and forms a reactive nuclear core maintained by feeding unused pebbles to the bed's top surface while used or burned-out pebbles run out and discharge through the outlet. A substantially conical body with its apex pointing upwardly and its periphery spaced from the periphery of the container spreads the bottom of the bed outwardly to provide an annular flow down the funnel-shaped bottom forming the runout, to the discharge outlet. This provides a largely constant downward velocity of the spheres throughout the diameter of the bed throughout a substantial portion of the down travel, so that all spheres reach about the same burned-out condition when they leave the core, after a single pass through the core area

  7. PEBBLES Simulation of Static Friction and New Static Friction Benchmark

    International Nuclear Information System (INIS)

    Cogliati, Joshua J.; Ougouag, Abderrafi M.

    2010-01-01

    Pebble bed reactors contain large numbers of spherical fuel elements arranged randomly. Determining the motion and location of these fuel elements is required for calculating certain parameters of pebble bed reactor operation. This paper documents the PEBBLES static friction model. This model uses a three dimensional differential static friction approximation extended from the two dimensional Cundall and Strack model. The derivation of determining the rotational transformation of pebble to pebble static friction force is provided. A new implementation for a differential rotation method for pebble to container static friction force has been created. Previous published methods are insufficient for pebble bed reactor geometries. A new analytical static friction benchmark is documented that can be used to verify key static friction simulation parameters. This benchmark is based on determining the exact pebble to pebble and pebble to container static friction coefficients required to maintain a stable five sphere pyramid.

  8. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    Science.gov (United States)

    Cisneros, Anselmo Tomas, Jr.

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

  9. Graphite for high-temperature reactors

    International Nuclear Information System (INIS)

    Hammer, W.; Leushacke, D.F.; Nickel, H.; Theymann, W.

    1976-01-01

    The different graphites necessary for HTRs are being developed, produced and tested within the Federal German ''Development Programme Nuclear Graphite''. Up to now, batches of the following graphite grades have been manufactured and fully characterized by the SIGRI Company to demonstrate reproducibility: pitch coke graphite AS2-500 for the hexagonal fuel elements and exchangeable reflector blocks; special pitch coke graphite ASI2-500 for reflector blocks of the pebble-bed reactor and as back-up material for the hexagonal fuel elements; graphite for core support columns. The material data obtained fulfill most of the requirements under present specifications. Production of large-size blocks for the permanent side reflector and the core support blocks is under way. The test programme covers all areas important for characterizing and judging HTR-graphites. In-pile testing comprises evaluation of the material for irradiation-induced changes of dimensions, mechanical and thermal properties - including behaviour under temperature cycling and creep behaviour - as well as irradiating fuel element segments and blocks. Testing out-of-pile includes: evaluation of corrosion rates and influence of corrosion on strength; strength measurements; including failure criteria. The test programme has been carried out extensively on the AS2-graphite, and the results obtained show that this graphite is suitable as HTGR fuel element graphite. (author)

  10. The effects of applying silicon carbide coating on core reactivity of pebble-bed HTR in water ingress accident

    Energy Technology Data Exchange (ETDEWEB)

    Zuhair, S.; Setiadipura, Topan [National Nuclear Energy Agency of Indonesia, Serpong Tagerang Selatan (Indonesia). Center for Nuclear Reactor Technology and Safety; Su' ud, Zaki [Bandung Institute of Technology (Indonesia). Dept. of Physics

    2017-03-15

    Graphite is used as the moderator, fuel barrier material, and core structure in High Temperature Reactors (HTRs). However, despite its good thermal and mechanical properties below the radiation and high temperatures, it cannot avoid corrosion as a consequence of an accident of water/air ingress. Degradation of graphite as a main HTR material and the formation of dangerous CO gas is a serious problem in HTR safety. One of the several steps that can be adopted to avoid or prevent the corrosion of graphite by the water/air ingress is the application of a thin layer of silicon carbide (SiC) on the surface of the fuel element. This study investigates the effect of applying SiC coating on the fuel surfaces of pebble-bed HTR in water ingress accident from the reactivity points of view. A series of reactivity calculations were done with the Monte Carlo transport code MCNPX and continuous energy nuclear data library ENDF/B-VII at temperature of 1200 K. Three options of UO{sub 2}, PuO{sub 2}, and ThO{sub 2}/UO{sub 2} fuel kernel were considered to obtain the inter comparison of the core reactivity of pebble-bed HTR in conditions of water/air ingress accident. The calculation results indicated that the UO{sub 2}-fueled pebble-bed HTR reactivity was slightly reduced and relatively more decreased when the thickness of the SiC coating increased. The reactivity characteristic of ThO{sub 2}/UO{sub 2}-fueled pebble-bed HTR showed a similar trend to that of UO{sub 2}, but did not show reactivity peak caused by water ingress. In contrast with UO{sub 2}- and ThO{sub 2}-fueled pebble-bed HTR, although the reactivity of PuO{sub 2}-fueled pebble-bed HTR was the lowest, its characteristics showed a very high reactivity peak (0.33 Δk/k) and this introduction of positive reactivity is difficult to control. SiC coating on the surface of the plutonium fuel pebble has no significant impact. From the comparison between reactivity characteristics of uranium, thorium and plutonium cores with 0

  11. Process for the production of prismatic graphite molded articles for high temperature fuel elements

    International Nuclear Information System (INIS)

    Huschka, H.; Rachor, L.; Hrovat, M.; Wolff, W.

    1976-01-01

    Prismatic graphite molded objects for high temperature fuel elements are prepared by producing the outer geometry and the holes for cooling channels and for receiving fuel and fertile materials in the formation of the carbon object

  12. Experimental performance and results of the critical pebble bed facility KAHTER

    Energy Technology Data Exchange (ETDEWEB)

    Krings, F. J.; Drueke, V.; Kirch, N.; Neef, R. D.

    1974-10-15

    The paper provides a description and results of critical experiments performed in KAHTER fueled with pebbles containing coated particles of HEU/Th oxide with a ratio of uranium-to-thorium of 1.1:5. Core loadings with varying amounts of fuel and solid graphite pebbles were tested with fuel-to-graphite pebble ratios of 3:1, 1:1, and 1:3. Tests included criticality for various fuel loadings with all control rods removed, control rod worths for reflector-mounted control as single rods and in a bank and control worths for a central control rod, reaction rates by flux wire activations (Dy, Mn, In, Au, and U-235) and detector measurements (BF3 and fission chamber), simulated xenon stability testing using the motions of a Cf-252 source and Cd-absorber observed by an externally-located BF3 detector, and the reactivity worth of a Hf burnable absorber. For calculations of the room-temperature zero-power critical experiment, the values for nitrogen and hydrogen contents of the graphite were taken from previous experiments in CESAR.

  13. Graphite fuels combustion off-gas treatment options

    International Nuclear Information System (INIS)

    Kirkham, R.J.; Lords, R.E.

    1993-03-01

    Scenarios for burning bulk graphite and for burning crushed fuel particles from graphite spent nuclear fuels have been considered. Particulates can be removed with sintered metal filters. Subsequent cooling would then condense semi-volatile fission products into or onto a particulate. These particulates would be trapped by a second sintered metal filter or downstream packed bed. A packed bed scrub column can be used to eliminate most of the iodine-129 and tritium. A molecular sieve bed is proposed to collect the residual 129 I and other tramp radionuclides downstream (Ruthenium, etc.). Krypton-85 can be recovered, if need be, either by cryogenics or by the KALC process (Krypton Adsorption in Liquid Carbon dioxide). Likewise carbon-14 in the form of carbon dioxide could be collected with a caustic or lime scrub solution and incorporated into a grout. Sulfur dioxide present will be well below regulatory concern level of 4.0 tons per year and most of it would be removed by the scrubber. Carbon monoxide emissions will depend on the choice of burner and start-up conditions. Should the system exceed the regulatory concern level, a catalytic converter in the final packed bed will be provided. Radon and its daughters have sufficiently short half-lives (less than two minutes). If necessary, an additional holdup bed can be added before the final HEPA filters or additional volume can be added to the molecular sieve bed to limit radon emissions. The calculated total effective dose equivalent at the Idaho National Engineering Laboratory boundary from a single release of all the 3 , 14 C, 85 Kr, and 129 I in the total fuel mass if 0.43 mrem/year

  14. 3D Nondestructive Visualization and Evaluation of TRISO Particles Distribution in HTGR Fuel Pebbles Using Cone-Beam Computed Tomography

    Directory of Open Access Journals (Sweden)

    Gongyi Yu

    2017-01-01

    Full Text Available A nonuniform distribution of tristructural isotropic (TRISO particles within a high-temperature gas-cooled reactor (HTGR pebble may lead to excessive thermal gradients and nonuniform thermal expansion during operation. If the particles are closely clustered, local hotspots may form, leading to excessive stresses on particle layers and an increased probability of particle failure. Although X-ray digital radiography (DR is currently used to evaluate the TRISO distributions in pebbles, X-ray DR projection images are two-dimensional in nature, which would potentially miss some details for 3D evaluation. This paper proposes a method of 3D visualization and evaluation of the TRISO distribution in HTGR pebbles using cone-beam computed tomography (CBCT: first, a pebble is scanned on our high-resolution CBCT, and 2D cross-sectional images are reconstructed; secondly, all cross-sectional images are restructured to form the 3D model of the pebble; then, volume rendering is applied to segment and display the TRISO particles in 3D for visualization and distribution evaluation. For method validation, several pebbles were scanned and the 3D distributions of the TRISO particles within the pebbles were produced. Experiment results show that the proposed method provides more 3D than DR, which will facilitate pebble fabrication research and production quality control.

  15. Use of plutonium and minor actinides as fuel in high temperature pebble bed reactors for waste minimization

    International Nuclear Information System (INIS)

    Meier, Astrid; Bernnat, Wolfgang; Lohnert, Guenther

    2009-01-01

    Energy production by nuclear fission gives rise to longlived radionuclides, such as plutonium and americium. The ''PuMA'' (Plutonium and Minor Actinides Waste Management) research project within the 6th Framework Program of the European Union serves to minimize waste arisings and transmute plutonium and minor actinides from spent LWR fuel elements by means of modular high-temperature reactors (HTR). Coating the fuel, which consists of kernels approx. 250 μm in radius and surrounded by graphite as the moderator material, allows very high operating and accident temperatures and very high burnups. One point examined is whether the inherent safety characteristics known for uranium oxide also exist for (PuO 2 + MAO 2 ) fuel. On the basis of a reference reactor similar to the South African PBMR-400, various loading strategies at maximum burnup are considered with a view to the inherent safety of the HTR. (orig.)

  16. Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores

    International Nuclear Information System (INIS)

    Stroh, K.R.

    1979-03-01

    The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases

  17. Thermogravimetric and Differential Scanning Calorimetric Behavior of Ball-Milled Nuclear Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eung Seon; Kim, Min Hwan; Kim, Yong Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Yi Hyun; Cho, Seung Yon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    An examination was made to characterize the oxidation behavior of ball-milled nuclear graphite powder through a TG-DSC analysis. With the ball milling time, the BET surface area increased with the reduction of particle size, but decreased with the chemisorptions of O{sub 2} on the activated surface. The enhancement of the oxidation after the ball milling is attributed to both increases in the specific surface area and atomic scale defects in the graphite structure. In a high temperature gas-cooled reactor, nuclear graphite has been widely used as fuel elements, moderator or reflector blocks, and core support structures owing to its excellent moderating power, mechanical properties and machinability. For the same reason, it will be used in a helium cooled ceramic reflector test blanket module for the ITER. Each submodule has a seven-layer breeding zone, including three neutron multiplier layers packed with beryllium pebbles, three lithium ceramic pebbles packed tritium breeder layers, and a reflector layer packed with 1 mm diameter graphite pebbles to reduce the volume of beryllium. The abrasion of graphite structures owing to relative motion or thermal cycle during operation may produce graphite dust. It is expected that graphite dust will be more oxidative than bulk graphite, and thus the oxidation behavior of graphite dust must be examined to analyze the safety of the reactors during an air ingress accident. In this study, the thermal stability of ball-milled graphite powder was investigated using a simultaneous thermogravimeter-differential scanning calorimeter.

  18. Fuel elements for high temperature reactors having special suitability for reuse of the structural graphite

    International Nuclear Information System (INIS)

    Huschka, H.; Herrmann, F.J.

    1976-01-01

    There are prepared fuel elements for high temperature reactors from which the fuel zone can be removed from the structural graphite after the burnup of the fissile material has taken place so that the fuel element can be filled with new fuel and again placed in the reactor by having the strength of the matrix in the fuel zone sufficient for binding the embedded coated fuel particles but substantially less than the strength of the structural graphite whereby by the action of force it can be easily split up without destroying the particles

  19. Studies on design principles and criteria of fuels and graphites for experimental multi-purpose very high temperature reactor

    International Nuclear Information System (INIS)

    Arai, Taketoshi; Sato, Sadao; Tani, Yutaro

    1977-12-01

    Design principles and criteria of fuels and graphites have been studied to determine the main design parameters of a reference core MARK-III of the Experimental Multi-purpose Very High Temperature Reactor. The present status of research and development for HTGR fuels and graphites is reviewed from a standpoint of their integrity and safety aspects, and is compared to the specific design requirements for the VHTR fuels and graphites. Consequently, reasonable materials specifications, safety criteria and design analysis methods are presented for coated fuel particle, fuel compact, graphite sleeve, core support graphite and neutron absorber material. These design principles and criteria will be refined by further experimental investigations. (auth.)

  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. Development of a system for rapid discharge of spherical fuel elements as a diversitary afterheat removal system for pebble-bed HTR-type reactors

    International Nuclear Information System (INIS)

    Phlippen, P.W.

    1982-07-01

    Owing to its spherical fuel elements the pebble bed high temperature reactor provides the possibility to remove these fuel elements rapidly from the reactor for the purpose of after-heat removal and cooling in situations of danger and to collect them in easily cooled tanks. The paper investigates and represents fields of problems such as critically behaviour of core and fuel element collecting tanks, emission time of the core, thermodynamics in the vessel etc. by example of the PNP-500 reactor concept. A selection for the construction proposal was made from the performance possibilities of the three necessary main components prestressed-concrete vessel closure, fuel element guide and fuel element storage with cooling system. The proposal includes a prestressed concrete vessel closure opening by hydraulics as well as three annular fuel element storage modules cooled with the containment air by natural convection. (orig.) [de

  2. Criteria for the selection of graphites for HTR integral block fuel elements

    International Nuclear Information System (INIS)

    Knowles, A.N.

    1980-01-01

    This paper is concerned with the special requirements for integral block fuel elements of the type first used in the Fort St. Vrain reactor. The main idea of these elements is that the carrier block and separate graphite clad fuel pins are combined into a single monolith. This combination leads to lower fabrication costs and some improvement in the thermal performance (lower temperature difference between fuel and the surface of heat transfer into the coolant). The advent of block fuel for HTRs of the Fort St. Vrain type has placed a fresh emphasis on the selection of graphite for block manufacture in respect of physical properties. This is because the temperature distributions typical of such fuelled blocks lead to shutdown stresses close to the maximum the graphite can sustain without damage. Figures presented in this paper suggest that the physical properties of the graphite can play a relatively large part in reducing such stress levels and that guidance on the key requirements for suitable specifications is therefore particularly needed by the manufacturers of fuel block graphites. While graphites for fuel blocks have this special need for combinations of physical properties which lead to low thermal and shrinkage stresses, the other characteristics must also receive attention. A low graphite cost combined with good homogeneity in the brick, so that waste minimized, are still necessary, while isotropy is also very important

  3. HTR-Proteus Pebble Bed Experimental Program Cores 5,6,7,&8: Columnar Hexagonal Point-on-Point Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

    Energy Technology Data Exchange (ETDEWEB)

    Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sterbentz, James W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Snoj, Luka [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lengar, Igor [Idaho National Lab. (INL), Idaho Falls, ID (United States); Koberl, Oliver [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.

  4. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 5, 6, 7, & 8: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:2 MODERATOR-TO-FUEL PEBBLE RATIO

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2013-03-01

    PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.

  5. The problem of gas gap between graphite - fuel channel reduction impact at Ignalina NPP

    International Nuclear Information System (INIS)

    1999-01-01

    Safety analysis of Ignalina NPP operation in the case when gap closure between graphite - fuel channel occur was performed. The main results of this analysis as well as data of gap measurements during the year 1996 - 1998 are provided

  6. Burnup analysis of a peu a peu fuel-loading scheme in a pebble bed reactor using the Monte Carlo method

    International Nuclear Information System (INIS)

    Irwanto, Dwi; Obara, Toru

    2010-01-01

    The design of a pebble bed reactor can be simplified by removing the unloading device from the system. For this reactor design, a suitable fuel-loading scheme is the peu a peu (little by little) fueling scheme. In the peu a peu mode, there is no unloading device; as such, the fuels are never discharged and remain at the bottom of the core during reactor operation. This means that the burnup cycle and reactivity is controlled by the addition of fuel. In this study, the Monte Carlo method is used to perform calculations with high accuracy. However, the calculation procedures for the peu a peu mode using the Monte Carlo method require lot of steps. Therefore, a computer code to automate the process of the peu a peu fuel-loading scheme based on the Monte Carlo MVP/MVP-BURN code has been developed using Fortran. Using the method developed in this study, burnup characteristics for a reference design of a small 20-MW pebble bed reactor with the peu a peu concept were analyzed. (author)

  7. Plutonium burning in a pebble-bed type high temperature nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bende, E.E

    2000-01-24

    in an HTR pebble. The Dancoff factors used in the pebble-cell calculations of chapters 3 and 4 were calculated according to this analytic expression. Appendix B gives the temperature profile across a pebble and derives, on the basis of this profile, the average and effective temperature of the fuel and the graphite. Appendix C presents a study on the maximisation of the Doppler-effect (i.e. the fuel temperature reactivity feedback). refs.

  8. Experimental investigation on feasibility of two-region-designed pebble-bed high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Yang Xingtuan; Hu Wenping; Jiang Shengyao

    2009-01-01

    Phenomenological experiments were performed on a 2-dimensional scaled model of the two-region designed pebble-bed high-temperature gas-cooled reactor core consisting of the distinct fuel pebble region and graphite pebble region. Issues with respect to the feasibility of the two-region design, including the establishment of the two-region arrangement, the mixing zone between the two regions, and the stagnant zone existence, were investigated. Three equilibrium conditions were proposed to evaluate the stable two-region arrangement formation. The general characteristics of the flow of the pebble bed were analyzed on basis of the observed phenomenon. It was found that a stable two-region arrangement was formed under the experimental conditions: the pebbles' motion was to some extent random but also confined by the neighbors of pebbles so that the mixing zone is constrained to a reasonable size. Guide plates utilized to improve mixing are proved to be effective without noticeable effect on the two-region arrangement features. Stagnant zones were observed under the experimental conditions and they were expected to be avoided by improving the design of the experimental setup. (author)

  9. Failure analysis of pebble bed reactors during earthquake by discrete element method

    International Nuclear Information System (INIS)

    Keppler, Istvan

    2013-01-01

    Highlights: ► We evaluated the load acting on the central reflector beam of a pebble bed reactor. ► The load acting on the reflector beam highly depends on fuel element distribution. ► The contact force values do not show high dependence on fuel element distribution. ► Earthquake increases the load of the reflector, not the contact forces. -- Abstract: Pebble bed reactors (PBR) are graphite-moderated, gas-cooled nuclear reactors. PBR reactors use a large number of spherical fuel elements called pebbles. From mechanical point of view, the arrangement of “small” spherical fuel elements in a container poses the same problem, as the so-called silo problem in powder technology and agricultural engineering. To get more exact information about the contact forces arising between the fuel elements in static and dynamic case, we simulated the static case and the effects of an earthquake on a model reactor by using discrete element method. We determined the maximal contact forces acting between the individual fuel elements. We found that the value of the maximal bending moment in the central reflector beam has a high deviation from the average value even in static case, and it can significantly increase in case of an earthquake. Our results can help the engineers working on the design of such types of reactors to get information about the contact forces, to determine the dust production and the crush probability of fuel elements within the reactor, and to model different accident scenarios

  10. Failure analysis of pebble bed reactors during earthquake by discrete element method

    Energy Technology Data Exchange (ETDEWEB)

    Keppler, Istvan, E-mail: keppler.istvan@gek.szie.hu [Department of Mechanics and Engineering Design, Szent István University, Páter K.u.1., Gödöllő H-2103 (Hungary)

    2013-05-15

    Highlights: ► We evaluated the load acting on the central reflector beam of a pebble bed reactor. ► The load acting on the reflector beam highly depends on fuel element distribution. ► The contact force values do not show high dependence on fuel element distribution. ► Earthquake increases the load of the reflector, not the contact forces. -- Abstract: Pebble bed reactors (PBR) are graphite-moderated, gas-cooled nuclear reactors. PBR reactors use a large number of spherical fuel elements called pebbles. From mechanical point of view, the arrangement of “small” spherical fuel elements in a container poses the same problem, as the so-called silo problem in powder technology and agricultural engineering. To get more exact information about the contact forces arising between the fuel elements in static and dynamic case, we simulated the static case and the effects of an earthquake on a model reactor by using discrete element method. We determined the maximal contact forces acting between the individual fuel elements. We found that the value of the maximal bending moment in the central reflector beam has a high deviation from the average value even in static case, and it can significantly increase in case of an earthquake. Our results can help the engineers working on the design of such types of reactors to get information about the contact forces, to determine the dust production and the crush probability of fuel elements within the reactor, and to model different accident scenarios.

  11. Quasi-direct numerical simulation of a pebble bed configuration. Part I: Flow (velocity) field analysis

    International Nuclear Information System (INIS)

    Shams, A.; Roelofs, F.; Komen, E.M.J.; Baglietto, E.

    2013-01-01

    Highlights: ► Quasi direct numerical simulations (q-DNS) of a pebble bed configuration has been performed. ► This q-DNS database may serve as a reference for the validation of different turbulence modeling approaches. ► A wide range of qualitative and quantitative data throughout the computational domain has been generated. ► Results for mean, RMS and covariance of velocity field are extensively reported in this paper. -- Abstract: High temperature reactors (HTR) are being considered for deployment around the world because of their excellent safety features. The fuel is embedded in a graphite moderator and can sustain very high temperatures. However, the appearance of hot spots in the pebble bed cores of HTR's may affect the integrity of the pebbles. A good prediction of the flow and heat transport in such a pebble bed core is a challenge for available turbulence models and such models need to be validated. In the present article, quasi direct numerical simulations (q-DNS) of a pebble bed configuration are reported, which may serve as a reference for the validation of different turbulence modeling approaches. Such approaches can be used in order to perform calculations for a randomly arranged pebble bed. Simulations are performed at a Reynolds number of 3088, based on pebble diameter, with a porosity level of 0.42. Detailed flow analyses have shown complex physics flow behavior and make this case challenging for turbulence model validation. Hence, a wide range of qualitative and quantitative data for velocity and temperature field have been extracted for this benchmark. In the present article (part I), results related to the flow field (mean, RMS and covariance of velocity) are documented and discussed in detail. Moreover, the discussion regarding the temperature field will be published in a separate article

  12. Reactivity considerations for the on-line refuelling of a pebble bed modular reactor—Illustrating safety for the most reactive core fuel load

    International Nuclear Information System (INIS)

    Reitsma, Frederik

    2012-01-01

    In the multi-pass fuel management scheme employed for the pebble bed modular reactor the fuel pebbles are re-circulated until they reach the target burn-up. The rate at which fresh fuel is loaded and burned fuel is discharged is a result of the core neutronics cycle analysis but in practice (on the plant) this has to be controlled and managed by the fuel handling and storage system and use of the burnup measurement system. The excess reactivity is the additional reactivity available in the core during operating conditions that is the result of loading a fuel mixture in the core that is more reactive (less burned) than what is required to keep the reactor critical at full power operational conditions. The excess reactivity is balanced by the insertion of the control rods to keep the reactor critical. The excess reactivity allows flexibility in operations, for example to overcome the xenon build up when power is decreased as part of load follow. In order to limit reactivity excursions and to ensure safe shutdown the excess reactivity and thus the insertion depth of the control rods at normal operating conditions has to be managed. One way to do this is by operational procedures. The reactivity effect of long-term operation with the control rods inserted deeper than the design point is investigated and a control rod insertion limit is proposed that will not limit normal operations. The effects of other phenomena that can increase the power defect, such as higher-than-expected fuel temperatures, are also introduced. All of these cases are then evaluated by ensuring cold shutdown is still achievable and where appropriate by reactivity insertion accident analysis. These aspects are investigated on the PBMR 400 MW design.

  13. Benchmark Evaluation of HTR-PROTEUS Pebble Bed Experimental Program

    International Nuclear Information System (INIS)

    Bess, John D.; Montierth, Leland; Köberl, Oliver

    2014-01-01

    Benchmark models were developed to evaluate 11 critical core configurations of the HTR-PROTEUS pebble bed experimental program. Various additional reactor physics measurements were performed as part of this program; currently only a total of 37 absorber rod worth measurements have been evaluated as acceptable benchmark experiments for Cores 4, 9, and 10. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the 235 U enrichment of the fuel, impurities in the moderator pebbles, and the density and impurity content of the radial reflector. Calculations of k eff with MCNP5 and ENDF/B-VII.0 neutron nuclear data are greater than the benchmark values but within 1% and also within the 3σ uncertainty, except for Core 4, which is the only randomly packed pebble configuration. Repeated calculations of k eff with MCNP6.1 and ENDF/B-VII.1 are lower than the benchmark values and within 1% (~3σ) except for Cores 5 and 9, which calculate lower than the benchmark eigenvalues within 4σ. The primary difference between the two nuclear data libraries is the adjustment of the absorption cross section of graphite. Simulations of the absorber rod worth measurements are within 3σ of the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments

  14. Pebble bed reactors simulation using MCNP: The Chinese HTR-10 reactor

    Directory of Open Access Journals (Sweden)

    SA Hosseini

    2013-09-01

    Full Text Available   Given the role of Gas-Graphite reactors as the fourth generation reactors and their recently renewed importance, in 2002 the IAEA proposed a set of Benchmarking problems. In this work, we propose a model both efficient in time and resources and exact to simulate the HTR-10 reactor using MCNP-4C code. During the present work, all of the pressing factors in PBM reactor design such as the inter-pebble leakage, fuel particle distribution and fuel pebble packing fraction effects have been taken into account to obtain an exact and easy to run model. Finally, the comparison between the results of the present work and other calculations made at INEEL proves the exactness of the proposed model.

  15. Effect of the Heat Treatment on the Graphite Matrix of Fuel Element for HTGR

    International Nuclear Information System (INIS)

    Lee, Chungyong; Lee, Seungjae; Suh, Jungmin; Jo, Youngho; Lee, Youngwoo; Cho, Moonsung

    2013-01-01

    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 for 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 Phenol as a binder were chosen and mixed with each other, formed and heated for the compressive strength test. The objective of this research is to optimize the kinds and composition of the mixed graphite and the forming process by evaluating the compressive strength before/after heat treatment (carbonization of binder). In this study, the effect of heat treatment on graphite matrix was studied in terms of the density and the compressive strength. The size (diameter and length) of pellet is increased by heat treatment. Due to additional weight reduction and swelling (length and diameter) of samples the density of graphite pellet is decreased from about 2.0 to about 1.7g/cm 3 . From the mechanical test results, the compressive strength of graphite pellets was related to the various conditions such as the contents of binder, the kinds of graphite and the heat treatment. Both the green pellet and the heat treated pellet, the compressive strength of G+S+P pellets is relatively higher than that of R+S+P pellets. To optimize fuel element matrix, the effect of Phenol and other binders, graphite composition and the heat treatment on the mechanical properties will be deeply investigated for further study

  16. A Preliminary Study on Calculation of Inter-Pebble Dancoff Factor in a Pebble Type Core

    International Nuclear Information System (INIS)

    Kim, Song Hyun; Kim, Hong Chul; Kim, Soon Young; Noh, Jae Man; Kim, Jong Kyung

    2009-01-01

    The Dancoff factor is an entering probability of the neutron escaped from specific fuel kernel to another one without the interaction with moderators. Currently, Dancoff factors are mainly evaluated from stochastic methods, hence a research on analytical method is considerably insufficient in this field. In order to analytically evaluate Dancoff factor considering double-heterogeneous effect, inter-pebble and intra-pebble Dancoff factors should be calculated, respectively. Intra-pebble Dancoff factor related with the fuel kernels in one pebble was analyzed in past study. For the evaluation of inter-pebble Dancoff factor, fuel region to region Dancoff factor (FRDF) was defined and the method to calculate the FRDF is developed in this study. The result is compared with the calculation result of the MCNP5 code

  17. Characterization of graphite dust produced by pneumatic lift

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Ke [Guangdong Provincial Key Laboratory of Thermal Management Engineering and Materials, Graduate School at Shenzhen, Tsinghua University, Shenzhen 518055, Guangdong (China); Peng, Wei; Liu, Bing [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Kang, Feiyu [Guangdong Provincial Key Laboratory of Thermal Management Engineering and Materials, Graduate School at Shenzhen, Tsinghua University, Shenzhen 518055, Guangdong (China); Yang, Xiaoyong; Li, Weihua [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Yu, Suyuan, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Educations, Tsinghua University, Beijing 100084 (China)

    2016-08-15

    Highlights: • Generation of graphite dust by pneumatic lift. • Determination of morphology and particle size distribution of graphite dust. • The size of graphite dust in this study is compared to AVR and THTR-300 results. • Graphite dust originates from both filler and binder of the matrix graphite. - Abstract: Graphite dust is an important safety concern of high-temperature gas-cooled reactor (HTR). The graphite dust could adsorb fission products, and the radioactive dust is transported by the coolant gas and deposited on the surface of the primary loop. The simulation of coagulation, aggregation, deposition, and resuspension behavior of graphite dust requires parameters such as particle size distribution and particle shape, but currently very limited data on graphite dust is available. The only data we have are from AVR and THTR-300, however, the AVR result is likely to be prejudiced by the oil ingress. In pebble-bed HTR, graphite dust is generally produced by mechanical abrasion, in particular, by the abrasion of graphite pebbles in the lifting pipe of the fuel handling system. Here we demonstrate the generation and characterization of graphite dust that were produced by pneumatic lift. This graphite dust could substitute the real dust in HTR for characterization. The dust, exhibiting a lamellar morphology, showed a number-weighted average particle size of 2.38 μm and a volume-weighted average size of 14.62 μm. These two sizes were larger than the AVR and THTR results. The discrepancy is possibly due to the irradiation effect and prejudice caused by the oil ingress accident. It is also confirmed by the Raman spectrum that both the filler particle and binder contribute to the dust generation.

  18. Nuclear graphite for high temperature reactors

    International Nuclear Information System (INIS)

    Marsden, B.J.

    2001-01-01

    The cores and reflectors in modern High Temperature Gas Cooled Reactors (HTRs) are constructed from graphite components. There are two main designs; the Pebble Bed design and the Prism design. In both of these designs the graphite not only acts as a moderator, but is also a major structural component that may provide channels for the fuel and coolant gas, channels for control and safety shut off devices and provide thermal and neutron shielding. In addition, graphite components may act as a heat sink or conduction path during reactor trips and transients. During reactor operation, many of the graphite component physical properties are significantly changed by irradiation. These changes lead to the generation of significant internal shrinkage stresses and thermal shut down stresses that could lead to component failure. In addition, if the graphite is irradiated to a very high irradiation dose, irradiation swelling can lead to a rapid reduction in modulus and strength, making the component friable.The irradiation behaviour of graphite is strongly dependent on its virgin microstructure, which is determined by the manufacturing route. Nevertheless, there are available, irradiation data on many obsolete graphites of known microstructures. There is also a well-developed physical understanding of the process of irradiation damage in graphite. This paper proposes a specification for graphite suitable for modern HTRs. (author)

  19. HTGR Fuels and Core Development Program. Quarterly progress report for the period ending August 31, 1977. [Graphite and fuel irradiation; fission product release

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    The work reported includes studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and data are presented.

  20. A methodology to investigate the contribution of conduction and radiation heat transfer to the effective thermal conductivity of packed graphite pebble beds, including the wall effect

    Energy Technology Data Exchange (ETDEWEB)

    De Beer, M., E-mail: maritz.db@gmail.com [School of Mechanical and Nuclear Engineering, North-West University, Private Bag X6001, Potchefstroom 2520 (South Africa); Du Toit, C.G., E-mail: Jat.DuToit@nwu.ac.za [School of Mechanical and Nuclear Engineering, North-West University, Private Bag X6001, Potchefstroom 2520 (South Africa); Rousseau, P.G., E-mail: pieter.rousseau@uct.ac.za [Department of Mechanical Engineering, University of Cape Town, Private Bag X3, Rondebosch 7701 (South Africa)

    2017-04-01

    Highlights: • The radiation and conduction components of the effective thermal conductivity are separated. • Near-wall effects have a notable influence on the effective thermal conductivity. • Effective thermal conductivity is a function of the macro temperature gradient. • The effective thermal conductivity profile shows a characteristic trend. • The trend is a result of the interplay between conduction and radiation. - Abstract: The effective thermal conductivity represents the overall heat transfer characteristics of a packed bed of spheres and must be considered in the analysis and design of pebble bed gas-cooled reactors. During depressurized loss of forced cooling conditions the dominant heat transfer mechanisms for the passive removal of decay heat are radiation and conduction. Predicting the value of the effective thermal conductivity is complex since it inter alia depends on the temperature level and temperature gradient through the bed, as well as the pebble packing structure. The effect of the altered packing structure in the wall region must therefore also be considered. Being able to separate the contributions of radiation and conduction allows a better understanding of the underlying phenomena and the characteristics of the resultant effective thermal conductivity. This paper introduces a purpose-designed test facility and accompanying methodology that combines physical measurements with Computational Fluid Dynamics (CFD) simulations to separate the contributions of radiation and conduction heat transfer, including the wall effects. Preliminary results obtained with the methodology offer important insights into the trends observed in the experimental results and provide a better understanding of the interplay between the underlying heat transfer phenomena.

  1. Influence of graphite discs, chamfers, and plenums on temperature distributions in high burnup fuel

    International Nuclear Information System (INIS)

    Ranger, A.; Tayal, M.; Singh, P.

    1990-04-01

    Previous studies have demonstrated the desirability to increase the fuel burnups in CANDU reactors from 7-9 GW.d/t to 21 GW.d/t. At high burnups, one consideration in fuel integrity is fission gas pressure, which is predicted to reach about 13 MPa. The gas pressure can be kept below the coolant pressure (about 10 MPa) via a variety of options such as bigger chamfers, deeper dishes, central hole, and plenums. However, it is important to address the temperature perturbations produced by the bigger chamfers and plenums which in turn, affect the gas pressure. Another consideration in fuel integrity is to reduce the likelihood of fuel failures via environmentally assisted cracking. Insertion of graphite discs between neighbouring pellets will lower the pellet temperatures, hence, lower fission gas release and lower expansion of the pellet. Therefore, it is desired to quantify the effect of graphite discs on pellet temperatures. Thermal analyses of different fuel element geometries: with and without chamfers, graphite discs, and plenums were performed. The results indicate that the two-dimensional distributions of temperatures due to the presence of chamfers, graphite discs, or plenums can have a significant impact on the integrity of high burnup fuel as we have been able to quantify in this paper

  2. Method of producing exfoliated graphite composite compositions for fuel cell flow field plates

    Energy Technology Data Exchange (ETDEWEB)

    Zhamu, Aruna; Shi, Jinjun; Guo, Jiusheng; Jang, Bor Z

    2014-04-08

    A method of producing an electrically conductive composite composition, which is particularly useful for fuel cell bipolar plate applications. The method comprises: (a) providing a supply of expandable graphite powder; (b) providing a supply of a non-expandable powder component comprising a binder or matrix material; (c) blending the expandable graphite with the non-expandable powder component to form a powder mixture wherein the non-expandable powder component is in the amount of between 3% and 60% by weight based on the total weight of the powder mixture; (d) exposing the powder mixture to a temperature sufficient for exfoliating the expandable graphite to obtain a compressible mixture comprising expanded graphite worms and the non-expandable component; (e) compressing the compressible mixture at a pressure within the range of from about 5 psi to about 50,000 psi in predetermined directions into predetermined forms of cohered graphite composite compact; and (f) treating the so-formed cohered graphite composite to activate the binder or matrix material thereby promoting adhesion within the compact to produce the desired composite composition. Preferably, the non-expandable powder component further comprises an isotropy-promoting agent such as non-expandable graphite particles. Further preferably, step (e) comprises compressing the mixture in at least two directions. The method leads to composite plates with exceptionally high thickness-direction electrical conductivity.

  3. Investigations on accidents with massive water ingress exemplified by the pebble bed reactor PNP-500

    International Nuclear Information System (INIS)

    Moormann, R.

    1986-01-01

    A computer code is used for analyses of massive water ingress accidents in the High-Temperature Gas Cooled Reactor concept PNP-500 with pebble bed core. The analyses are mainly focussed on graphite corrosion processes. For the investigated accidents a correct reactor shut down in assumed. The mass of water ingressing into the primary circuit is varied between 1000 and 7500 kg (i.e., up to hypothetical values). The dependence of accident consequences on parameters such as intensity and starting time of the afterheat removal system or kinetic values of the chemical processes is examined. The results show that even under pessimistic assumptions the extent of the graphite corrosion is relatively low; significant damaging of fuel elements or graphite components does not occur. A primary circuit depressurization, combined with local burning of water gas, would probably not affect the fission product retention potential of the (gastight) containment. Summing up, the risk caused by these accidents remains small. (orig.) [de

  4. On the Evaluation of Pebble Bead Reactor Critical Experiments Using the Pebbed Code

    International Nuclear Information System (INIS)

    Gougar, Hans D.; Sen, R. Sonat

    2014-01-01

    Critical experiments pose a particular but necessary challenge to validating pebble bed reactor design codes. Fuel and core heterogeneities, impurities in graphite, variable packing of pebbles, and moderately strong neutronic coupling are among the factors that inject uncertainty into the results obtained with lower fidelity core physics models. Some of these are addressed in this study. The PEBBED pebble bed reactor fuel management code under development at the Idaho National Laboratory is designed for rapid design and analysis of pebble bed high temperature reactors (PBRs). Embedded within the code are the THERMIX-KONVEK thermal fluid solver and the COMBINE-7 spectrum generation code for inline cross section homogenization. Because 1D symmetry can be found at each stage of core heterogeneity; spherical at TRISO and pebble levels, and cylindrical at the control rod and core levels, the 1-D transport capability of ANISN is assumed to be sufficient in most cases for generating flux solutions for cross section homogenization. Furthermore, it is fast enough to be executed during the analysis or the equilibrium core. Multi-group diffusion-based design codes such as PEBBED and VSOP are not expected to yield the accuracy and resolution of continuous energy Monte Carlo codes for evaluation of critical experiments. Nonetheless, if the preparation of multigroup cross sections can adequately capture the physics of the mixing of PBR fuel elements and leakage from the core, reasonable results may be obtained. In this paper, results of the application of PEBBED to two critical experiments (HTR Proteus and HTR-10) and associated computational models are presented. The embedded 1-D transport solver is shown to capture the double heterogeneity of the pebble fuel in unit cell calculations. Eigenvalue calculations of a whole core are more challenging, particularly if the boron concentration is uncertain. The sensitivity of major safety parameters to variations in modeling

  5. A Comparison of Materials Issues for Cermet and Graphite-Based NTP Fuels

    Science.gov (United States)

    Stewart, Mark E.; Schnitzler, Bruce G.

    2013-01-01

    This paper compares material issues for cermet and graphite fuel elements. In particular, two issues in NTP fuel element performance are considered here: ductile to brittle transition in relation to crack propagation, and orificing individual coolant channels in fuel elements. Their relevance to fuel element performance is supported by considering material properties, experimental data, and results from multidisciplinary fluid/thermal/structural simulations. Ductile to brittle transition results in a fuel element region prone to brittle fracture under stress, while outside this region, stresses lead to deformation and resilience under stress. Poor coolant distribution between fuel element channels can increase stresses in certain channels. NERVA fuel element experimental results are consistent with this interpretation. An understanding of these mechanisms will help interpret fuel element testing results.

  6. Temperature Analysis and Failure Probability of the Fuel Element in HTR-PM

    International Nuclear Information System (INIS)

    Yang Lin; Liu Bing; Tang Chunhe

    2014-01-01

    Spherical fuel element is applied in the 200-MW High Temperature Reactor-Pebble-bed Modular (HTR-PM). Each spherical fuel element contains approximately 12,000 coated fuel particles in the inner graphite matrix with a diameter of 50mm to form the fuel zone, while the outer shell with a thickness of 5mm is a fuel-free zone made up of the same graphite material. Under high burnup irradiation, the temperature of fuel element rises and the stress will result in the damage of fuel element. The purpose of this study is to analyze the temperature of fuel element and to discuss the stress and failure probability. (author)

  7. Fission Product Transport and Source Terms in HTRs: Experience from AVR Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    Rainer Moormann

    2008-01-01

    Full Text Available Fission products deposited in the coolant circuit outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (DBA. The deposited fission products may be released in depressurization accidents because present pebble bed HTR concepts abstain from a gas tight containment. Contamination of the circuit also hinders maintenance work. Experiments, performed from 1972 to 88 on the AVR, an experimental pebble bed HTR, allow for a deeper insight into fission product transport behavior. The activity deposition per coolant pass was lower than expected and was influenced by fission product chemistry and by presence of carbonaceous dust. The latter lead also to inconsistencies between Cs plate out experiments in laboratory and in AVR. The deposition behavior of Ag was in line with present models. Dust as activity carrier is of safety relevance because of its mobility and of its sorption capability for fission products. All metal surfaces in pebble bed reactors were covered by a carbonaceous dust layer. Dust in AVR was produced by abrasion in amounts of about 5 kg/y. Additional dust sources in AVR were ours oil ingress and peeling of fuel element surfaces due to an air ingress. Dust has a size of about 1  m, consists mainly of graphite, is partly remobilized by flow perturbations, and deposits with time constants of 1 to 2 hours. In future reactors, an efficient filtering via a gas tight containment is required because accidents with fast depressurizations induce dust mobilization. Enhanced core temperatures in normal operation as in AVR and broken fuel pebbles have to be considered, as inflammable dust concentrations in the gas phase.

  8. Neutronic design of a Liquid Salt-cooled Pebble Bed Reactor (LSPBR)

    International Nuclear Information System (INIS)

    De Zwaan, S. J.; Boer, B.; Lathouwers, D.; Kloosterman, J. L.

    2006-01-01

    A renewed interest has been raised for liquid salt cooled nuclear reactors. The excellent heat transfer properties of liquid salt coolants provide several benefits, like lower fuel temperatures, higher coolant outlet temperatures, increased core power density and better decay heat removal. In order to benefit from the online refueling capability of a pebble bed reactor, the Liquid Salt Pebble Bed Reactor (LSPBR) is proposed. This is a high temperature pebble-bed reactor with a fuel design similar to existing HTRs, but using a liquid salt as a coolant. In this paper, the selection criteria for the liquid salt coolant are described. Based on its neutronic properties, LiF-BeF 2 (FLIBE) was selected for the LSPBR. Two designs of the LSPBR were considered: a cylindrical core and an annular core with a graphite inner reflector. Coupled neutronic-thermal hydraulic calculations were performed to obtain the steady state power distribution and the corresponding fuel temperatures. Finally, calculations were performed to investigate the decay heat removal capability in a protected loss-of-forced cooling accident. The maximum allowable power that can be produced with the LSPBR is hereby determined. (authors)

  9. Multidisciplinary Simulation of Graphite-Composite and Cermet Fuel Elements for NTP Point of Departure Designs

    Science.gov (United States)

    Stewart, Mark E.; Schnitzler, Bruce G.

    2015-01-01

    This paper compares the expected performance of two Nuclear Thermal Propulsion fuel types. High fidelity, fluid/thermal/structural + neutronic simulations help predict the performance of graphite-composite and cermet fuel types from point of departure engine designs from the Nuclear Thermal Propulsion project. Materials and nuclear reactivity issues are reviewed for each fuel type. Thermal/structural simulations predict thermal stresses in the fuel and thermal expansion mis-match stresses in the coatings. Fluid/thermal/structural/neutronic simulations provide predictions for full fuel elements. Although NTP engines will utilize many existing chemical engine components and technologies, nuclear fuel elements are a less developed engine component and introduce design uncertainty. Consequently, these fuel element simulations provide important insights into NTP engine performance.

  10. Calculation of thermal stresses in graphite fuel blocks

    International Nuclear Information System (INIS)

    Lejeail, Y.; Cabrillat, M.T.

    2005-01-01

    This paper presents a parametric study of temperature and thermal stress calculations inside a HTGR core graphite block, taking into account the effect of fluence on the thermal and mechanical properties, up to 4. 10 21 n/cm 2 . The Finite Element model, realized with Cast3M CEA code, includes the effects of irradiation creep, which tends to produce secondary stress relaxation. Then, the Weibull weakest link theory is recalled, evaluating the possible effects of volume, stress field distribution (loading factor), and multiaxiality for graphite-type materials, and giving the methodology to compare the stress to rupture for the structure to the one obtained from characterization, in the general case. The maximum of the Weibull stress in Finite Element calculations is compared to the value for tensile specimens. It is found that the maximum of the stress corresponds to the end of the irradiation cycle, after reactor shutdown, since both thermal conductivity and Young's modulus increase with time. However, this behaviour is partly counterbalanced by the increase of material strength with irradiation. (authors)

  11. Fabrication of uranium carbide/beryllium carbide/graphite experimental-fuel-element specimens

    International Nuclear Information System (INIS)

    Muenzer, W.A.

    1978-01-01

    A method has been developed for fabricating uranium carbide/beryllium carbide/graphite fuel-element specimens for reactor-core-meltdown studies. The method involves milling and blending the raw materials and densifying the resulting blend by conventional graphite-die hot-pressing techniques. It can be used to fabricate specimens with good physical integrity and material dispersion, with densities of greater than 90% of the theoretical density, and with a uranium carbide particle size of less than 10 μm

  12. Non-catalyzed cathodic oxygen reduction at graphite granules in microbial fuel cells

    International Nuclear Information System (INIS)

    Freguia, Stefano; Rabaey, Korneel; Yuan Zhiguo; Keller, Juerg

    2007-01-01

    Oxygen is the most sustainable electron acceptor currently available for microbial fuel cell (MFC) cathodes. However, its high overpotential for reduction to water limits the current that can be produced. Several materials and catalysts have previously been investigated in order to facilitate oxygen reduction at the cathode surface. This study shows that significant stable currents can be delivered by using a non-catalyzed cathode made of granular graphite. Power outputs up to 21 W m -3 (cathode total volume) or 50 W m -3 (cathode liquid volume) were attained in a continuous MFC fed with acetate. These values are higher than those obtained in several other studies using catalyzed graphite in various forms. The presence of nanoscale pores on granular graphite provides a high surface area for oxygen reduction. The current generated with this cathode can sustain an anodic volume specific COD removal rate of 1.46 kg COD m -3 d -1 , which is higher than that of a conventional aerobic process. This study demonstrates that microbial fuel cells can be operated efficiently using high surface graphite as cathode material. This implies that research on microbial fuel cell cathodes should not only focus on catalysts, but also on high surface area materials

  13. Non-catalyzed cathodic oxygen reduction at graphite granules in microbial fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Freguia, Stefano; Rabaey, Korneel; Yuan, Zhiguo; Keller, Juerg [The University of Queensland, St. Lucia, Qld (Australia). Advanced Wastewater Management Centre

    2007-12-01

    Oxygen is the most sustainable electron acceptor currently available for microbial fuel cell (MFC) cathodes. However, its high overpotential for reduction to water limits the current that can be produced. Several materials and catalysts have previously been investigated in order to facilitate oxygen reduction at the cathode surface. This study shows that significant stable currents can be delivered by using a non-catalyzed cathode made of granular graphite. Power outputs up to 21 W m{sup -3} (cathode total volume) or 50 W m{sup -3} (cathode liquid volume) were attained in a continuous MFC fed with acetate. These values are higher than those obtained in several other studies using catalyzed graphite in various forms. The presence of nanoscale pores on granular graphite provides a high surface area for oxygen reduction. The current generated with this cathode can sustain an anodic volume specific COD removal rate of 1.46 kg{sub COD} m{sup -3} d{sup -1}, which is higher than that of a conventional aerobic process. This study demonstrates that microbial fuel cells can be operated efficiently using high surface graphite as cathode material. This implies that research on microbial fuel cell cathodes should not only focus on catalysts, but also on high surface area materials. (author)

  14. The Cross-Flow Mixing Analysis of Quasi-Static Pebble Flow in Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Fang Xiang; Liu Zhiyong; Sun Yanfei; Yang Xingtuan; Jiang Shengyao

    2014-01-01

    In the pebble bed reactor, large number of fuel pebbles’ movement law and moving state can affect the reactor’s design, operation and safety directly. Therefore the pebble flow, which is based on the theory of particle streaming, is one of the most important research subjects of the pebble bed reactor engineering. The in-core pebble flow is a very slow particle flow (or called quasi-static particle flow), which is very different from the usual particle motion. How to accurately describe the characteristics of in-core pebble flow is a central issue for this subject. Due to the presence of random flow, the cross-mixing phenomenon will occur inevitably. In the present paper, the mixing phenomenon of pebble flow is generalized on the basis of experiment results. The pebble flow cross-mixing probability serves as the parameter which describes both the regularity and the randomness of pebble flow. The results are provided in the form of diagrammatic presentation. (author)

  15. Anisotropic Material Behavior of Uni-axially Compacted Graphite Matrix for HTGR Fuel Compact Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Yoon, Ji-Hae; Cho, Moon Sung [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In developing the fuel compact fabrication technology, and fuel graphite material to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions and the material properties of fuel element. It was observed, during this development, that the pressing technique employed for the compaction fabrication prior to the two successive heat treatments (carbonization and final high temperature heat treatment) was of extreme importance in determining the material properties of the final compact product. In this work, the material behavior of the uni-axially pressed graphite matrix during the carbonization and final heat treatment are evaluated and summarized along the different directions, viz., perpendicular and parallel directions to pressing direction. In this work, the dimensional variations and variations in thermal expansion, thermal conductivity and Vickers hardness of the graphite matrix compact samples in the axial and radial directions prepared by uni-axial pressing are evaluated, and compared with those of samples prepared by cold isostatic pressing with the available data. From this work, the followings are observed. 1) Dimensional changes of matrix graphite green compacts during carbonization show that the difference in radial and axial variations shows a large anisotropic behavior in shrinkage. The radial variation is very small while the axial variation is large. During carbonization, the stresses caused by the force would be released in to the axial direction together with the phenolic resin vapor. 2) Dimensional variation of compact samples in perpendicular and parallel directions during carbonization shows a large difference in behavior when compact sample is prepared by uni-axial pressing. However, when compact sample is prepared by cold isostatic pressing, there is

  16. Anisotropic Material Behavior of Uni-axially Compacted Graphite Matrix for HTGR Fuel Compact Fabrication

    International Nuclear Information System (INIS)

    Lee, Young-Woo; Yeo, Seunghwan; Yoon, Ji-Hae; Cho, Moon Sung

    2016-01-01

    In developing the fuel compact fabrication technology, and fuel graphite material to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions and the material properties of fuel element. It was observed, during this development, that the pressing technique employed for the compaction fabrication prior to the two successive heat treatments (carbonization and final high temperature heat treatment) was of extreme importance in determining the material properties of the final compact product. In this work, the material behavior of the uni-axially pressed graphite matrix during the carbonization and final heat treatment are evaluated and summarized along the different directions, viz., perpendicular and parallel directions to pressing direction. In this work, the dimensional variations and variations in thermal expansion, thermal conductivity and Vickers hardness of the graphite matrix compact samples in the axial and radial directions prepared by uni-axial pressing are evaluated, and compared with those of samples prepared by cold isostatic pressing with the available data. From this work, the followings are observed. 1) Dimensional changes of matrix graphite green compacts during carbonization show that the difference in radial and axial variations shows a large anisotropic behavior in shrinkage. The radial variation is very small while the axial variation is large. During carbonization, the stresses caused by the force would be released in to the axial direction together with the phenolic resin vapor. 2) Dimensional variation of compact samples in perpendicular and parallel directions during carbonization shows a large difference in behavior when compact sample is prepared by uni-axial pressing. However, when compact sample is prepared by cold isostatic pressing, there is

  17. Kr-85m activity as burnup measurement indicator in a pebble bed reactor based on ORIGEN2.1 Computer Simulation

    Science.gov (United States)

    Husnayani, I.; Udiyani, P. M.; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    Pebble Bed Reactor (PBR) is a high temperature gas-cooled reactor which employs graphite as a moderator and helium as a coolant. In a multi-pass PBR, burnup of the fuel pebble must be measured in each cycle by online measurement in order to determine whether the fuel pebble should be reloaded into the core for another cycle or moved out of the core into spent fuel storage. One of the well-known methods for measuring burnup is based on the activity of radionuclide decay inside the fuel pebble. In this work, the activity and gamma emission of Kr-85m were studied in order to investigate the feasibility of Kr-85m as burnup measurement indicator in a PBR. The activity and gamma emission of Kr-85 were estimated using ORIGEN2.1 computer code. The parameters of HTR-10 were taken as a case study in performing ORIGEN2.1 simulation. The results show that the activity revolution of Kr-85m has a good relationship with the burnup of the pebble fuel in each cycle. The Kr-85m activity reduction in each burnup step,in the range of 12% to 4%, is considered sufficient to show the burnup level in each cycle. The gamma emission of Kr-85m is also sufficiently high which is in the order of 1010 photon/second. From these results, it can be concluded that Kr-85m is suitable to be used as burnup measurement indicator in a pebble bed reactor.

  18. Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

    International Nuclear Information System (INIS)

    Karahan, Aydın; Kazimi, Mujid S.

    2013-01-01

    The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors

  19. Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydın, E-mail: karahan@alum.mit.edu; Kazimi, Mujid S.

    2013-10-15

    The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

  20. 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

  1. Study on thermal conductivity of HTR spherical fuel element matrix graphite

    International Nuclear Information System (INIS)

    Zhang Kaihong; Liu Xiaoxue; Zhao Hongsheng; Li Ziqiang; Tang Chunhe

    2014-01-01

    Taking the spherical fuel element matrix graphite ball samples as an example, this paper introduced the principle and method of laser thermal conductivity meter, as well as the specific heat capacity, and analyzed the effects of different test methods and sampling methods on the thermal conductivities at 1000 ℃ of graphite material. The experimental results show that the thermal conductivities of graphite materials tested by synchronous thermal analyzer combining with laser thermal conductivity meter were different from that directly by laser thermal conductivity meter, the former was more reliable and accurate than the later; When sampling from different positions, central samples had higher thermal conductivities than edging samples, which was related to the material density and porosity at the different locations; the thermal conductivities had obvious distinction between samples from different directions, which was because the layer structure of polycrystalline graphite preferred orientation under pressure, generally speaking, the thermal conductivities perpendicular to the molding direction were higher than that parallel to the molding direction. Besides this, the test results show that the thermal conductivities of all the graphite material samples were greater than 30 W/(m (K), achieving the thermal performance index of high temperature gas cooled reactor. (authors)

  2. Dimensional Behavior of Matrix Graphite Compacts during Heat Treatments for HTGR Fuel Element Fabrication

    International Nuclear Information System (INIS)

    Lee, Young-Woo; Yeo, Seunghwan; Cho, Moon Sung

    2015-01-01

    The carbonization is a process step where the binder that is incorporated during the matrix graphite powder preparation step is evaporated and the residue of the binder is carbonized during the heat treatment at about 1073 K. This carbonization step is followed by the final high temperature heat treatment where the carbonized compacts are heat treated at 2073-2173 K in vacuum for a relatively short time (about 2 hrs). In order to develop a fuel compact fabrication technology, and for fuel matrix graphite to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions, which has a strong influence on the further steps and the material properties of fuel element. In this work, the dimensional changes of green compacts during the carbonization and final heat treatment are evaluated when compacts have different densities from different pressing conditions and different final heat treatment temperatures are employed, keeping other process parameters constant, such as the binder content, carbonization time, temperature and atmosphere (two hours ant 1073K and N2 atmosphere). In this work, the dimensional variations of green compacts during the carbonization and final heat treatment are evaluated when compacts have different densities from different pressing conditions and different final heat treatment temperatures are employed

  3. Modelling of HTR (High Temperature Reactor Pebble-Bed 10 MW to Determine Criticality as A Variations of Enrichment and Radius of the Fuel (Kernel With the Monte Carlo Code MCNP4C

    Directory of Open Access Journals (Sweden)

    Hammam Oktajianto

    2014-12-01

    Full Text Available Gas-cooled nuclear reactor is a Generation IV reactor which has been receiving significant attention due to many desired characteristics such as inherent safety, modularity, relatively low cost, short construction period, and easy financing. High temperature reactor (HTR pebble-bed as one of type of gas-cooled reactor concept is getting attention. In HTR pebble-bed design, radius and enrichment of the fuel kernel are the key parameter that can be chosen freely to determine the desired value of criticality. This paper models HTR pebble-bed 10 MW and determines an effective of enrichment and radius of the fuel (Kernel to get criticality value of reactor. The TRISO particle coated fuel particle which was modelled explicitly and distributed in the fuelled region of the fuel pebbles using a Simple-Cubic (SC lattice. The pebble-bed balls and moderator balls distributed in the core zone using a Body-Centred Cubic lattice with assumption of a fresh fuel by the fuel enrichment was 7-17% at 1% range and the size of the fuel radius was 175-300 µm at 25 µm ranges. The geometrical model of the full reactor is obtained by using lattice and universe facilities provided by MCNP4C. The details of model are discussed with necessary simplifications. Criticality calculations were conducted by Monte Carlo transport code MCNP4C and continuous energy nuclear data library ENDF/B-VI. From calculation results can be concluded that an effective of enrichment and radius of fuel (Kernel to achieve a critical condition was the enrichment of 15-17% at a radius of 200 µm, the enrichment of 13-17% at a radius of 225 µm, the enrichments of 12-15% at radius of 250 µm, the enrichments of 11-14% at a radius of 275 µm and the enrichment of 10-13% at a radius of 300 µm, so that the effective of enrichments and radii of fuel (Kernel can be considered in the HTR 10 MW. Keywords—MCNP4C, HTR, enrichment, radius, criticality 

  4. Experimental and computational investigation of flow of pebbles in a pebble bed nuclear reactor

    Science.gov (United States)

    Khane, Vaibhav B.

    The Pebble Bed Reactor (PBR) is a 4th generation nuclear reactor which is conceptually similar to moving bed reactors used in the chemical and petrochemical industries. In a PBR core, nuclear fuel in the form of pebbles moves slowly under the influence of gravity. Due to the dynamic nature of the core, a thorough understanding about slow and dense granular flow of pebbles is required from both a reactor safety and performance evaluation point of view. In this dissertation, a new integrated experimental and computational study of granular flow in a PBR has been performed. Continuous pebble re-circulation experimental set-up, mimicking flow of pebbles in a PBR, is designed and developed. Experimental investigation of the flow of pebbles in a mimicked test reactor was carried out for the first time using non-invasive radioactive particle tracking (RPT) and residence time distribution (RTD) techniques to measure the pebble trajectory, velocity, overall/zonal residence times, flow patterns etc. The tracer trajectory length and overall/zonal residence time is found to increase with change in pebble's initial seeding position from the center towards the wall of the test reactor. Overall and zonal average velocities of pebbles are found to decrease from the center towards the wall. Discrete element method (DEM) based simulations of test reactor geometry were also carried out using commercial code EDEM(TM) and simulation results were validated using the obtained benchmark experimental data. In addition, EDEM(TM) based parametric sensitivity study of interaction properties was carried out which suggests that static friction characteristics play an important role from a packed/pebble beds structural characterization point of view. To make the RPT technique viable for practical applications and to enhance its accuracy, a novel and dynamic technique for RPT calibration was designed and developed. Preliminary feasibility results suggest that it can be implemented as a non

  5. 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.

  6. PEM fuel cells with injection moulded bipolar plates of highly filled graphite compounds; PEM-Brennstoffzellen mit spritzgegossenen Bipolarplatten aus hochgefuelltem Graphit-Compound

    Energy Technology Data Exchange (ETDEWEB)

    Kreuz, Can

    2008-04-11

    This work concerns with the injection moulding of highly filled graphite compounds to bipolar plates for PEM fuel cells in a power output range between 100 - 500 Watts. A particular focus is laid on the combination of the three multidisciplinary scopes like material development, production technology and component development / design. The results of the work are specified by the process-oriented characterisation of the developed and manufactured bipolar plates as well as their application in a functioning fuel cell. (orig.)

  7. Calculation of the packing fraction in a pebble-bed ADS and redesigning of the Transmutation Advanced Device for Sustainable Energy Applications (TADSEA)

    International Nuclear Information System (INIS)

    García, L.; Pérez, J.; García, C.; Escrivá, A.; Rosales, J.; Abánades, A.

    2012-01-01

    Highlights: ► We based our study on an ADS for TRU transmutation and high temperature production. ► We calculated the number of pebbles that fit in a cylindrical ADS core. ► In both ADS design options studied, the mass of Pu isotopes reduces considerably. ► The system can reach coolant outlet temperatures high enough for hydrogen production. ► The maximum temperature values obtained in the ADS are not dangerous for TRISO fuel. - Abstract: One of the main problems that should be addressed in the use of nuclear fuels for heat and electricity production is the management of nuclear waste from conventional nuclear power plants and its inventory minimization. Fast reactors and Accelerator Driven Systems (ADSs) are the main options for reducing the long-lived radioactive waste inventory. In previous studies, the conceptual design of a Transmutation Advanced Device for Sustainable Energy Applications (TADSEA) has been made. The TADSEA is a pebble-bed ADS cooled by helium and moderated by graphite; it uses as fuel small amounts of transuranic elements in the form of TRISO particles, confined in 3 cm radius graphite pebbles. It has been conceived for Plutonium (Pu) and Minor Actinides (MA) transmutation and for achieving very high helium temperatures at the core's outlet to match the thermal requirements for hydrogen production by high temperature electrolysis (HTE) or by the iodine-sulfur (I–S) thermo-chemical cycle. In this paper, a geometrical method for calculating the real number of pebbles that fit in a cylindrical ADS core, according to its size and pebble configuration, is described. Based on its results, the packing fraction influence on the TADSEA's main work parameters is studied, and the redesign of the previous configuration is done in order to maintain the exit thermal power established in the preliminary design. Results have shown the capability of the system to reach coolant outlet temperatures high enough for its application to hydrogen

  8. Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Luther, Erik; Rooyen, Isabella van; Leckie, Rafael; Papin, Pallas; Nelson, Andrew; Hunter, James

    2015-03-01

    In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabrication must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.

  9. Effects of random pebble distribution on the multiplication factor in HTR pebble bed reactors

    Energy Technology Data Exchange (ETDEWEB)

    Auwerda, G.J., E-mail: g.j.auwerda@tudelft.n [Department of Physics of Nuclear Reactors at the Delft University of Technology, Mekelweg 15, Delft (Netherlands); Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der [Department of Physics of Nuclear Reactors at the Delft University of Technology, Mekelweg 15, Delft (Netherlands)

    2010-08-15

    In pebble bed reactors the pebbles have a random distribution within the core. The usual approach in modeling the bed is homogenizing the entire bed. To quantify the errors arising in such a model, this article investigates the effect on k{sub eff} of three phenomena in random pebble distributions: non-uniform packing density, neutron streaming in between the pebbles, and variations in Dancoff factor. For a 100 cm high cylinder with reflective top and bottom boundary conditions 25 pebble beds were generated. Of each bed three core models were made: a homogeneous model, a zones model including density fluctuations, and an exact model with all pebbles modeled individually. The same was done for a model of the PROTEUS facility. k{sub eff} calculations were performed with three codes: Monte Carlo, diffusion, and finite element transport. By comparing k{sub eff} of the homogenized and zones model the effect of including density fluctuations in the pebble bed was found to increase k{sub eff} by 71 pcm for the infinite cylinder and 649 pcm for PROTEUS. The large value for PROTEUS is due to the low packing fraction near the top of the pebble bed, causing a significant lower packing fraction for the bulk of the pebble bed in the homogenized model. The effect of neutron streaming was calculated by comparing the zones model with the exact model, and was found to decrease k{sub eff} by 606 pcm for the infinite cylinder, and by 1240 pcm for PROTEUS. This was compared with the effect of using a streaming correction factor on the diffusion coefficient in the zones model, which resulted in {Delta}{sub streaming} values of 340 and 1085 pcm. From this we conclude neutron streaming is an important effect in pebble bed reactors, and is not accurately described by the correction factor on the diffusion coefficient. Changing the Dancoff factor in the outer part of the pebble bed to compensate for the lower probability of neutrons to enter other fuel pebbles caused no significant changes

  10. Computational and experimental prediction of dust production in pebble bed reactors, Part II

    Energy Technology Data Exchange (ETDEWEB)

    Hiruta, Mie; Johnson, Gannon [Department of Mechanical Engineering, University of Idaho, 1776 Science Center Drive, Idaho Falls, ID 83401 (United States); Rostamian, Maziar, E-mail: mrostamian@asme.org [Department of Mechanical Engineering, University of Idaho, 1776 Science Center Drive, Idaho Falls, ID 83401 (United States); Potirniche, Gabriel P. [Department of Mechanical Engineering, University of Idaho, 1776 Science Center Drive, Idaho Falls, ID 83401 (United States); Ougouag, Abderrafi M. [Idaho National Laboratory, 2525 N Fremont Avenue, Idaho Falls, ID 83401 (United States); Bertino, Massimo; Franzel, Louis [Department of Physics, Virginia Commonwealth University, Richmond, VA 23284 (United States); Tokuhiro, Akira [Department of Mechanical Engineering, University of Idaho, 1776 Science Center Drive, Idaho Falls, ID 83401 (United States)

    2013-10-15

    Highlights: • Custom-built high temperature, high pressure tribometer is designed. • Two different wear phenomena at high temperatures are observed. • Experimental wear results for graphite are presented. • The graphite wear dust production in a typical Pebble Bed Reactor is predicted. -- Abstract: This paper is the continuation of Part I, which describes the high temperature and high pressure helium environment wear tests of graphite–graphite in frictional contact. In the present work, it has been attempted to simulate a Pebble Bed Reactor core environment as compared to Part I. The experimental apparatus, which is a custom-designed tribometer, is capable of performing wear tests at PBR relevant higher temperatures and pressures under a helium environment. This environment facilitates prediction of wear mass loss of graphite as dust particulates from the pebble bed. The experimental results of high temperature helium environment are used to anticipate the amount of wear mass produced in a pebble bed nuclear reactor.

  11. Application of boron and gadolinium burnable poison particles in UO2 and PUO2 fuels in HTRs

    International Nuclear Information System (INIS)

    Kloosterman, J.L.

    2003-01-01

    Burnup calculations have been performed on a standard HTR fuel pebble (fuel zone with radius of 2.5 cm surrounded with a 0.5 cm thick graphite layer) and burnable poison particles (BPPs) containing B 4 C made of pure 10 B or containing Gd 2 O 3 made of natural Gd. Two types of fuel were considered: UO 2 fuel made of 8% enriched uranium and PuO 2 fuel made of plutonium from LWR spent fuel. The radius of the BPP and the number of particles per fuel pebble were varied to find the flattest reactivity-to-time curve. For the UO 2 fuel, the reactivity swing is lowest (around 2%) for BPPs made of B 4 C with radius of 75 μm. In this case around 1070 BPPs per fuel pebble are needed. For the PuO 2 fuel to get a reactivity swing below 4%, the optimal radius of the BPP is the same, but the number of particles per fuel pebble should be around 1600. The optimal radius of the Gd 2 O 3 particles in the UO 2 fuel is about 10 times that of the B 4 C particles. The reactivity swing is around 3% when each fuel pebble contains only 9 BPPs with radius of 840 μm. The results of the Gd particles illustrate nicely the usage of black burnable poison particles introduced by Van Dam [Ann. Nuclear Energy 27 (2000) 733

  12. Laminated exfoliated graphite composite-metal compositions for fuel cell flow field plate or bipolar plate applications

    Science.gov (United States)

    Zhamu, Aruna; Shi, Jinjun; Guo, Jiusheng; Jang, Bor Z

    2014-05-20

    An electrically conductive laminate composition for fuel cell flow field plate or bipolar plate applications. The laminate composition comprises at least a thin metal sheet having two opposed exterior surfaces and a first exfoliated graphite composite sheet bonded to the first of the two exterior surfaces of the metal sheet wherein the exfoliated graphite composite sheet comprises: (a) expanded or exfoliated graphite and (b) a binder or matrix material to bond the expanded graphite for forming a cohered sheet, wherein the binder or matrix material is between 3% and 60% by weight based on the total weight of the first exfoliated graphite composite sheet. Preferably, the first exfoliated graphite composite sheet further comprises particles of non-expandable graphite or carbon in the amount of between 3% and 60% by weight based on the total weight of the non-expandable particles and the expanded graphite. Further preferably, the laminate comprises a second exfoliated graphite composite sheet bonded to the second surface of the metal sheet to form a three-layer laminate. Surface flow channels and other desired geometric features can be built onto the exterior surfaces of the laminate to form a flow field plate or bipolar plate. The resulting laminate has an exceptionally high thickness-direction conductivity and excellent resistance to gas permeation.

  13. Sana experiments for self-acting removal of the after-heat in reactors with pebble bed fuel and their interpretation

    International Nuclear Information System (INIS)

    Niessen, H.F.; Stoecker, Bernd; Amoignon, Olivier; Zuying, Gao; Jie, Liu

    1997-01-01

    For the confirmation of self-acting afterheat removal under hypothetical accident conditions from pebble bed reactors at the Research Center Juelich a test facility with an electrical heating input up to 30kW was erected and operated. A description of the test facility is given. Within the different tests the pebble diameter, the pebble material, the gas in the pebble bed, the heating-power and the arrangement of the heating were changed. Parts of the data were used within an IAEA Co-ordinated Research Program as benchmark problems for the code validation. All computer codes could simulate the test results with a sufficient good agreement, when the tests were executed with helium. For the tests with nitrogen the natural convection has to be taken into account. (author)

  14. Safeguards Challenges for Pebble-Bed Reactors (PBRs):Peoples Republic of China (PRC)

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, Charles W. [Massachusetts Institute of Technology (MIT); Moses, David Lewis [ORNL

    2009-11-01

    The Peoples Republic of China (PRC) is operating the HTR-10 pebble-bed reactor (PBR) and is in the process of building a prototype PBR plant with two modular reactors (250-MW(t) per reactor) feeding steam to a single turbine-generator. It is likely to be the first modular hightemperature reactor to be ready for commercial deployment in the world because it is a highpriority project for the PRC. The plant design features multiple modular reactors feeding steam to a single turbine generator where the number of modules determines the plant output. The design and commercialization strategy are based on PRC strengths: (1) a rapidly growing electric market that will support low-cost mass production of modular reactor units and (2) a balance of plant system based on economics of scale that uses the same mass-produced turbine-generator systems used in PRC coal plants. If successful, in addition to supplying the PRC market, this strategy could enable China to be the leading exporter of nuclear reactors to developing countries. The modular characteristics of the reactor match much of the need elsewhere in the world. PBRs have major safety advantages and a radically different fuel. The fuel, not the plant systems, is the primary safety system to prevent and mitigate the release of radionuclides under accident conditions. The fuel consists of small (6-cm) pebbles (spheres) containing coatedparticle fuel in a graphitized carbon matrix. The fuel loading per pebble is small (~9 grams of low-enriched uranium) and hundreds of thousands of pebbles are required to fuel a nuclear plant. The uranium concentration in the fuel is an order of magnitude less than in traditional nuclear fuels. These characteristics make the fuel significantly less attractive for illicit use (weapons production or dirty bomb); but, its unusual physical form may require changes in the tools used for safeguards. This report describes PBRs, what is different, and the safeguards challenges. A series of

  15. Safeguards Challenges for Pebble-Bed Reactors (PBRs):Peoples Republic of China (PRC)

    International Nuclear Information System (INIS)

    Forsberg, Charles W.; Moses, David Lewis

    2009-01-01

    The Peoples Republic of China (PRC) is operating the HTR-10 pebble-bed reactor (PBR) and is in the process of building a prototype PBR plant with two modular reactors (250-MW(t) per reactor) feeding steam to a single turbine-generator. It is likely to be the first modular high temperature reactor to be ready for commercial deployment in the world because it is a high priority project for the PRC. The plant design features multiple modular reactors feeding steam to a single turbine generator where the number of modules determines the plant output. The design and commercialization strategy are based on PRC strengths: (1) a rapidly growing electric market that will support low-cost mass production of modular reactor units and (2) a balance of plant system based on economics of scale that uses the same mass-produced turbine-generator systems used in PRC coal plants. If successful, in addition to supplying the PRC market, this strategy could enable China to be the leading exporter of nuclear reactors to developing countries. The modular characteristics of the reactor match much of the need elsewhere in the world. PBRs have major safety advantages and a radically different fuel. The fuel, not the plant systems, is the primary safety system to prevent and mitigate the release of radionuclides under accident conditions. The fuel consists of small (6-cm) pebbles (spheres) containing coated particle fuel in a graphitized carbon matrix. The fuel loading per pebble is small (∼9 grams of low-enriched uranium) and hundreds of thousands of pebbles are required to fuel a nuclear plant. The uranium concentration in the fuel is an order of magnitude less than in traditional nuclear fuels. These characteristics make the fuel significantly less attractive for illicit use (weapons production or dirty bomb); but, its unusual physical form may require changes in the tools used for safeguards. This report describes PBRs, what is different, and the safeguards challenges. A series of

  16. Modeling stationary and moving pebbles in a pebble bed reactor

    International Nuclear Information System (INIS)

    Zhao, Xiang; Montgomery, Trent; Zhang, Sijun

    2015-01-01

    Highlights: • The stationary and moving pebbles in a PBR are numerically studied by DEM. • The packing structure of stationary pebbles is simulated by a filling process. • The packing structural properties are obtained and analyzed. • The dynamic behavior of pebbles is predicted and discussed. - Abstract: This paper presents a numerical study of the stationary and moving pebbles in a pebble bed reactor (PBR) by means of discrete element method (DEM). The packing structure of stationary pebbles is simulated by a filling process that terminates with the settling of the pebbles into a PBR. The packing structural properties are obtained and analyzed. Subsequently, when the outlet of the PBR is opened during the operation of the PBR, the stationary pebbles start to flow downward and are removed at the bottom of the PBR. The dynamic behavior of pebbles is predicted and discussed. Our results indicate the DEM can offer both macroscopic and microscopic information for PBR design calculations and safety assessment

  17. Matrix formulation of pebble circulation in the pebbed code

    International Nuclear Information System (INIS)

    Gougar, H.D.; Terry, W.K.; Ougouag, A.M.

    2002-01-01

    The PEBBED technique provides a foundation for equilibrium fuel cycle analysis and optimization in pebble-bed cores in which the fuel elements are continuously flowing and, if desired, recirculating. In addition to the modern analysis techniques used in or being developed for the code, PEBBED incorporates a novel nuclide-mixing algorithm that allows for sophisticated recirculation patterns using a matrix generated from basic core parameters. Derived from a simple partitioning of the pebble flow, the elements of the recirculation matrix are used to compute the spatially averaged density of each nuclide at the entry plane from the nuclide densities of pebbles emerging from the discharge conus. The order of the recirculation matrix is a function of the flexibility and sophistication of the fuel handling mechanism. This formulation for coupling pebble flow and neutronics enables core design and fuel cycle optimization to be performed by the manipulation of a few key core parameters. The formulation is amenable to modern optimization techniques. (author)

  18. Comparison of the thorium- and low-enriched uranium fuel cycle in the OTTO pebble bed HTR

    Energy Technology Data Exchange (ETDEWEB)

    Teuchert, E; Maly, V

    1973-03-15

    From the study of the physical, technical and economical properties of the LOTTO and TOTTO fuel cycle the authors draw the conclusion that no fundamental reason can be found for a definite preference of one of the two cycles. Either of them can be developed to become a very attractive concept. The flexibility of this reactor allows the design performance in that way that the reactor becomes a safe system with reasonable economy. The decision for the preference of the LOTTO or TOTTO can be allowed to be governed by the requirements of the outer fuel cycle or by politics.

  19. Graphitic Carbon Nitride as a Catalyst Support in Fuel Cells and Electrolyzers

    International Nuclear Information System (INIS)

    Mansor, Noramalina; Miller, Thomas S.; Dedigama, Ishanka; Jorge, Ana Belen; Jia, Jingjing; Brázdová, Veronika; Mattevi, Cecilia; Gibbs, Chris; Hodgson, David; Shearing, Paul R.; Howard, Christopher A.; Corà, Furio; Shaffer, Milo; Brett, Daniel J.L.

    2016-01-01

    Highlights: • Graphitic carbon nitride (gCN) describes many materials with different structures. • gCNs can exhibit excellent mechanical, chemical and thermal resistance. • A major obstacle for pure gCN catalyst supports is limited electronic conductivity. • Composite/Hybrid gCN structures show excellent performance as catalyst supports. • gCNs have great potential for use in fuel calls and water electrolyzers. - Abstract: Electrochemical power sources, such as polymer electrolyte membrane fuel cells (PEMFCs), require the use of precious metal catalysts which are deposited as nanoparticles onto supports in order to minimize their mass loading and therefore cost. State-of-the-art/commercial supports are based on forms of carbon black. However, carbon supports present disadvantages including corrosion in the operating fuel cell environment and loss of catalyst activity. Here we review recent work examining the potential of different varieties of graphitic carbon nitride (gCN) as catalyst supports, highlighting their likely benefits, as well as the challenges associated with their implementation. The performance of gCN and hybrid gCN-carbon materials as PEMFC electrodes is discussed, as well as their potential for use in alkaline systems and water electrolyzers. We illustrate the discussion with examples taken from our own recent studies.

  20. Modeling stationary and dynamic pebbles in a pebble bed reactor

    International Nuclear Information System (INIS)

    Zhao, Xiang; Montgomery, Trent; Zhang, Sijun

    2011-01-01

    This paper presents a numerical study of the stationary and dynamic pebbles in a pebble bed reactor (PBR) by means of discrete element method (DEM). At first, the packing structure of stationary pebbles is simulated by filling process until the settling of pebbles into PBR. The packing structural properties are obtained and analyzed. Subsequently, when the outlet of PBR is open during the operational maintenance of PBR, the stationary pebbles start to flow downward and are removed at the bottom of PBR. The dynamic behavior of pebbles is predicted and discussed. Our results indicate the DEM can offer both macroscopic and microscopic information for PBR design calculations and safety assessment. (author)

  1. In situ formation of graphene layers on graphite surfaces for efficient anodes of microbial fuel cells.

    Science.gov (United States)

    Tang, Jiahuan; Chen, Shanshan; Yuan, Yong; Cai, Xixi; Zhou, Shungui

    2015-09-15

    Graphene can be used to improve the performance of the anode in a microbial fuel cell (MFC) due to its good biocompatibility, high electrical conductivity and large surface area. However, the chemical production and modification of the graphene on the anode are environmentally hazardous because of the use of various harmful chemicals. This study reports a novel method based on the electrochemical exfoliation of a graphite plate (GP) for the in situ formation of graphene layers on the surface of a graphite electrode. When the resultant graphene-layer-based graphite plate electrode (GL/GP) was used as an anode in an MFC, a maximum power density of 0.67 ± 0.034 W/m(2) was achieved. This value corresponds to 1.72-, 1.56- and 1.26-times the maximum power densities of the original GP, exfoliated-graphene-modified GP (EG/GP) and chemically-reduced-graphene-modified GP (rGO/GP) anodes, respectively. Electrochemical measurements revealed that the high performance of the GL/GP anode was attributable to its macroporous structure, improved electron transfer and high electrochemical capacitance. The results demonstrated that the proposed method is a facile and environmentally friendly synthesis technique for the fabrication of high-performance graphene-based electrodes for use in microbial energy harvesting. Copyright © 2015 Elsevier B.V. All rights reserved.

  2. Oxygen reduction kinetics on graphite cathodes in sediment microbial fuel cells.

    Science.gov (United States)

    Renslow, Ryan; Donovan, Conrad; Shim, Matthew; Babauta, Jerome; Nannapaneni, Srilekha; Schenk, James; Beyenal, Haluk

    2011-12-28

    Sediment microbial fuel cells (SMFCs) have been used as renewable power sources for sensors in fresh and ocean waters. Organic compounds at the anode drive anodic reactions, while oxygen drives cathodic reactions. An understanding of oxygen reduction kinetics and the factors that determine graphite cathode performance is needed to predict cathodic current and potential losses, and eventually to estimate the power production of SMFCs. Our goals were to (1) experimentally quantify the dependence of oxygen reduction kinetics on temperature, electrode potential, and dissolved oxygen concentration for the graphite cathodes of SMFCs and (2) develop a mechanistic model. To accomplish this, we monitored current on polarized cathodes in river and ocean SMFCs. We found that (1) after oxygen reduction is initiated, the current density is linearly dependent on polarization potential for both SMFC types; (2) current density magnitude increases linearly with temperature in river SMFCs but remains constant with temperature in ocean SMFCs; (3) the standard heterogeneous rate constant controls the current density temperature dependence; (4) river and ocean SMFC graphite cathodes have large potential losses, estimated by the model to be 470 mV and 614 mV, respectively; and (5) the electrochemical potential available at the cathode is the primary factor controlling reduction kinetic rates. The mechanistic model based on thermodynamic and electrochemical principles successfully fit and predicted the data. The data, experimental system, and model can be used in future studies to guide SMFC design and deployment, assess SMFC current production, test cathode material performance, and predict cathode contamination.

  3. FSV experience in support of the GT-MHR reactor physics, fuel performance, and graphite

    International Nuclear Information System (INIS)

    Baxter, A.M.; McEachern, D.; Hanson, D.L.; Vollman, R.E.

    1994-11-01

    The Fort St. Vrain (FSV) power plant was the most recent operating graphite-moderated, helium-cooled nuclear power plant in the United States. Many similarities exist between the FSV design and the current design of the GT-MHR. Both designs use graphite as the basic building blocks of the core, as structural material, in the reflectors, and as a neutron moderator. Both designs use hexagonal fuel elements containing cylindrical fuel rods with coated fuel particles. Helium is the coolant and the power densities vary by less than 5%. Since material and geometric properties of the GT-MHR core am very similar to the FSV core, it is logical to draw upon the FSV experience in support of the GT-MHR design. In the Physics area, testing at FSV during the first three cycles of operation has confirmed that the calculational models used for the core design were very successful in predicting the core nuclear performance from initial cold criticality through power operation and refueling. There was excellent agreement between predicted and measured initial core criticality and control rod positions during startup. Measured axial flux distributions were within 5% of the predicted value at the peak. The isothermal temperature coefficient at zero power was in agreement within 3%, and even the calculated temperature defect over the whole operating range for cycle 3 was within 8% of the measured defect. In the Fuel Performance area, fuel particle coating performance, and fission gas release predictions and an overall plateout analysis were performed for decommissioning purposes. A comparison between predicted and measured fission gas release histories of Kr-85m and Xe-138 and a similar comparison with specific circulator plateout data indicated good agreement between prediction and measured data. Only I-131 plateout data was overpredicted, while Cs-137 data was underpredicted

  4. MIT pebble bed reactor project

    Energy Technology Data Exchange (ETDEWEB)

    Kadak, Andrew C. [Massachusetts Institute of Technology, Cambridge (United States)

    2007-03-15

    The conceptual design of the MIT modular pebble bed reactor is described. This reactor plant is a 250 Mwth, 120 Mwe indirect cycle plant that is designed to be deployed in the near term using demonstrated helium system components. The primary system is a conventional pebble bed reactor with a dynamic central column with an outlet temperature of 900 C providing helium to an intermediate helium to helium heat exchanger (IHX). The outlet of the IHX is input to a three shaft horizontal Brayton Cycle power conversion system. The design constraint used in sizing the plant is based on a factory modularity principle which allows the plant to be assembled 'Lego' style instead of constructed piece by piece. This principle employs space frames which contain the power conversion system that permits the Lego-like modules to be shipped by truck or train to sites. This paper also describes the research that has been conducted at MIT since 1998 on fuel modeling, silver leakage from coated fuel particles, dynamic simulation, MCNP reactor physics modeling and air ingress analysis.

  5. MIT pebble bed reactor project

    International Nuclear Information System (INIS)

    Kadak, Andrew C.

    2007-01-01

    The conceptual design of the MIT modular pebble bed reactor is described. This reactor plant is a 250 Mwth, 120 Mwe indirect cycle plant that is designed to be deployed in the near term using demonstrated helium system components. The primary system is a conventional pebble bed reactor with a dynamic central column with an outlet temperature of 900 C providing helium to an intermediate helium to helium heat exchanger (IHX). The outlet of the IHX is input to a three shaft horizontal Brayton Cycle power conversion system. The design constraint used in sizing the plant is based on a factory modularity principle which allows the plant to be assembled 'Lego' style instead of constructed piece by piece. This principle employs space frames which contain the power conversion system that permits the Lego-like modules to be shipped by truck or train to sites. This paper also describes the research that has been conducted at MIT since 1998 on fuel modeling, silver leakage from coated fuel particles, dynamic simulation, MCNP reactor physics modeling and air ingress analysis

  6. Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled high temperature reactor

    International Nuclear Information System (INIS)

    Zhu, G.; Zou, Y.; Xu, H.

    2016-01-01

    Sustainability of thorium fuel in a Pebble-Bed Fluoride salt-cooled High temperature Reactor (PBFHR) is investigated to find the feasible region of high discharge burnup and negative Flibe (2LiF-BeF_2) salt Temperature Reactivity Coefficient (TRC). Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tri-structural-isotropic (TRISO) coated particle system for increasing fuel loading and decreasing excessive moderation. To analyze the neutronic characteristics, an equilibrium calculation method of thorium fuel self-sustainability is developed. We have compared two refueling schemes (mixing flow pattern and directional flow pattern) and two kinds of reflector materials (SiC and graphite). This method found that the feasible region of breeding and negative Flibe TRC is between 20 vol% and 62 vol% fuel loading in the fuel. A discharge burnup could be achieved up to about 200 MWd/kgHM. The case with directional flow pattern and SiC reflector showed superior burnup characteristics but the worst radial power peak factor, while the case with mixing flow pattern and SiC reflector, which was the best tradeoff between discharge burnup and radial power peak factor, could provide burnup of 140 MWd/kgHM and about 1.4 radial power peak factor with 50 vol% dispersion fuel. In addition, Flibe salt displays good neutron properties as a coolant of quasi-fast reactors due to the strong "9Be(n,2n) reaction and low neutron absorption of "6Li (even at 1000 ppm) in fast spectrum. Preliminary thermal hydraulic calculation shows a good safety margin. The greatest challenge of this reactor may be the decades irradiation time of the pebble fuel. (A.C)

  7. Pebble breakage in gravel

    International Nuclear Information System (INIS)

    Tuitz, C.

    2012-01-01

    The spatial clustering of broken pebbles in gravel layers of a Miocene sedimentary succession was investigated. Field observations suggested that the occurrence of broken pebbles could be related with gravel hosted shear deformation bands, which were the result of extensional regional deformation. Several different methods were used in this work to elucidate these observations. These methods include basic field work, measurements of physical pebble and gravel properties and, the application of different numerical modelling schemes. In particular, the finite element method in 2D and the discrete element method in 2D and 3D were used in order to quantify mechanisms of pebble deformation. The main objective of this work was to identify potential mechanisms that control particle breakage in fluvial gravel, which could explain the clustered spatial distribution of broken pebbles. The results of 2D finite element stress analysis indicated that the breakage load of differently located and oriented diametrical loading axes on a pebble varies and, that the weakest loading configuration coincides with the smallest principal axis of the pebble. The 3D discrete element method was applied to study the contact load distribution on pebbles in gravel deposits and the influence of different degrees of particle imbrication and orientation. The results showed that an increase of the number of imbricated particles leads to a significant load transfer from the rim to the centre of the oblate sides of the ellipsoidal particles. The findings of these pebble-scale investigations provided the basis for outcropscale modelling, where simulated gravel layers were subjected to layer-parallel extension. These outcrop-scale models revealed the existence of a particle breakage enhancing mechanism that becomes active during early stages of shear band formation. The interaction of such shear bands with the less deformed host material results in particle stress concentrations and subsequently

  8. Method of making a graphite fuel element having carbonaceous fuel bodies

    International Nuclear Information System (INIS)

    Miertschin, G.N.; Leary, D.F.

    1977-01-01

    Particulate nuclear fuel material, particulate carbon and pitch are combined with an additive which is effective to reduce the coke yield upon carbonization to mold a green fuel body. The additive may be polystyrene, a styrene-butadiene copolymer, an aromatic hydrocarbon having a molecular weight between about 75 and 300 or a saturated hydrocarbon polymer. The green fuel body is inserted in a complementary cavity within a porous nuclear fuel element body and heated in situ to decompose the pitch and additive, leaving a relatively close-fitting fuel body in the cavity

  9. 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

  10. Sockets and Pebbles

    Science.gov (United States)

    1997-01-01

    This close-up Sojourner rover image of a small rock shows that weathering has etched-out pebbles to produce sockets. In the image, sunlight is coming from the upper left. Sockets (with shadows on top) are visible at the lower left and pebbles (with bright tops and shadowed bases) are seen at the lower center and lower right. Two pebbles (about 0.5 cm across) are visible at the lower center.Mars Pathfinder is the second in NASA's Discovery program of low-cost spacecraft with highly focused science goals. The Jet Propulsion Laboratory, Pasadena, CA, developed and manages the Mars Pathfinder mission for NASA's Office of Space Science, Washington, D.C. JPL is a division of the California Institute of Technology (Caltech).

  11. Studies on air ingress for pebble bed reactors

    International Nuclear Information System (INIS)

    Moore, R.L.; Oh, C.H.; Merrill, B.J.; Petti, D.A.

    2002-01-01

    A loss-of-coolant accident (LOCA) has been considered a critical event for helium-cooled pebbled bed reactors. Following helium depressurization, it is anticipated that unless countermeasures are taken air will enter the core through the break and then by molecular diffusion and ultimately by natural convection leading to oxidation of the in-core graphite structure and graphite pebbles. Thus, without any mitigating features a LOCA will lead to an air ingress event. The INEEL is studying such an event with two well-respected light water reactor transient response codes: RELAP5/ATHENA and MELCOR. To study the degree of graphite oxidation occurring due to an air ingress event, a MELCOR model of a reference pebble bed design was constructed. A modified version of MELCOR developed at INEEL, which includes graphite oxidation capabilities, and molecular diffusion of air into helium was used for these calculations. Results show that the lower reflector graphite consumes all of the oxygen before reaching the core. The results also show a long time delay between the time that the depressurization phase of the accident is over and the time that natural circulation air through the core occurs. (author)

  12. 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)

  13. PEBBLES: A COMPUTER CODE FOR MODELING PACKING, FLOW AND RECIRCULATIONOF PEBBLES IN A PEBBLE BED REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2006-10-01

    A comprehensive, high fidelity model for pebble flow has been developed and embodied in the PEBBLES computer code. In this paper, a description of the physical artifacts included in the model is presented and some results from using the computer code for predicting the features of pebble flow and packing in a realistic pebble bed reactor design are shown. The sensitivity of models to various physical parameters is also discussed.

  14. The Performance of a Direct Borohydride/Peroxide Fuel Cell Using Graphite Felts as Electrodes

    Directory of Open Access Journals (Sweden)

    Heng-Yi Lee

    2017-08-01

    Full Text Available A direct borohydride/peroxide fuel cell (DBPFC generates electrical power by recirculating liquid anolyte and catholyte between the stack and reservoirs, which is similar to the operation of flow batteries. To enhance the accessibility of the catalyst layer to the liquid anolyte/catholyte, graphite felts are employed as the porous diffusion layer of a single-cell DBPFC instead of carbon paper/cloth. The effects of the type of anode alkaline solution and operating conditions, including flow rate and temperature of the anolyte/catholyte, on DBPFC performance are investigated and discussed. The durability of the DBPFC is also evaluated by galvanostatic discharge at 0.1 A∙cm−2 for over 50 h. The results of this preliminary study show that a DBPFC with porous graphite electrodes can provide a maximum power density of 0.24 W∙cm−2 at 0.8 V. The performance of the DBPFC drops slightly after 50 h of operation; however, the discharge capacity shows no significant decrease.

  15. Distribution of 60Co and 54Mn in graphite material of irradiated HTGR fuel assemblies

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Kikuchi, Teruo; Kobayashi, Fumiaki; Minato, Kazuo; Fukuda, Kousaku; Ikawa, Katsuichi; Iwamoto, Kazumi

    1984-05-01

    Distribution of 60 Co and 54 Mn was measured in the graphite sleeves and blocks of the third and fourth HTGR fuel assemblies irradiated in the Oarai Gas Loop-1 (OGL-1), which is a high temperature inpile gas loop installed in the Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Research Institute (JAERI). Axial and circumferential profiles were obtained by gamma spectrometry, and radial profiles by lathe sectioning with gamma spectrometry. Distribution of 60 Co is in good agreement with that of thermal neutron flux, and the Co content in the graphite is estimated to be -- 1 x 10 -9 in weight fraction. Concentration of 54 Mn decreases toward the axial center in its axial profile, and radially is almost uniform inside and appreciably higher at free surfaces. An estimated Fe content of --10 -8 in wight fraction is smaller by two orders of magnitude than that from chemical analysis. Higher concentraion of 60 Co and 54 Mn at the free surfaces suggests the importance of transportation process of these nuclides in the coolant loop. (author)

  16. CFD investigating the air ingress accident for a HTGR simulation of graphite corrosion oxidation

    International Nuclear Information System (INIS)

    Ferng, Y.M.; Chi, C.W.

    2012-01-01

    Highlights: ► A CFD model is proposed to investigate graphite oxidation corrosion in the HTR-10. ► A postulated air ingress accident is assumed in this paper. ► Air ingress flowrate is the predicted result, instead of the preset one. ► O 2 would react with graphite on pebble surface, causing the graphite corrosion. ► No fuel exposure is predicted to be occurred under the air ingress accident. - Abstract: Through a compressible multi-component CFD model, this paper investigates the characteristics of graphite oxidation corrosion in the HTR-10 core under the postulated accident of gas duct rupture. In this accident, air in the steam generator cavity would enter into the core after pressure equilibrium is achieved between the core and the cavity, which is also called as the air ingress accident. Oxygen in the air would react with graphite on pebble surface, subsequently resulting in oxidation corrosion and challenging fuel integrity. In this paper, characteristics of graphite oxidation corrosion during the air ingress accident can be reasonably captured, including distributions of graphite corrosion amount on the different cross-sections, time histories of local corrosion amount at the monitoring points and overall corrosion amount in the core, respectively. Based on the transient simulation results, the corrosion pattern and its corrosion rate would approach to the steady-state conditions as the accident continuously progresses. The total amount of graphite corrosion during a 3-day accident time is predicted to be about 31 kg with the predicted asymptotic corrosion rate. This predicted value is less than that from the previous work of Gao and Shi.

  17. Pebble Puzzle Solved

    Science.gov (United States)

    2004-01-01

    [figure removed for brevity, see original site] Figure 1 In the quest to determine if a pebble was jamming the rock abrasion tool on NASA's Mars Exploration Rover Opportunity, scientists and engineers examined this up-close, approximate true-color image of the tool. The picture was taken by the rover's panoramic camera, using filters centered at 601, 535, and 482 nanometers, at 12:47 local solar time on sol 200 (August 16, 2004). Colored spots have been drawn on this image corresponding to regions where panoramic camera reflectance spectra were acquired (see chart in Figure 1). Those regions are: the grinding wheel heads (yellow); the rock abrasion tool magnets (green); the supposed pebble (red); a sunlit portion of the aluminum rock abrasion tool housing (purple); and a shadowed portion of the rock abrasion tool housing (brown). These spectra demonstrated that the composition of the supposed pebble was clearly different from that of the sunlit and shadowed portions of the rock abrasion tool, while similar to that of the dust-coated rock abrasion tool magnets and grinding heads. This led the team to conclude that the object disabling the rock abrasion tool was indeed a martian pebble.

  18. A Graphite Oxide Paper Polymer Electrolyte for Direct Methanol Fuel Cells

    Directory of Open Access Journals (Sweden)

    Ravi Kumar

    2011-01-01

    Full Text Available A flow directed assembly of graphite oxide solution was used in the formation of free-standing graphene oxide paper of approximate thickness of 100 μm. The GO papers were characterised by XRD and SEM. Electrochemical characterization of the GO paper membrane electrode assembly revealed proton conductivities of 4.1 × 10−2 S cm−1 to 8.2 × 10−2 S cm−1 at temperatures of 25–90°C. A direct methanol fuel cell, at 60°C, gave a peak power density of 8 mW cm−2 at a current density of 35 mA cm−2.

  19. Preliminary neutronic study on Pu-based OTTO cycle pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Setiadipura, Topan; Zuhair [National Nuclear Energy Agency of Indonesia (BATAN), Selatan (Indonesia). Center for Nuclear Reactor Technology and Safety; Irwanto, Dwi [Bandung Institute of Technology (ITB), Bandung (Indonesia). Nuclear Physics and Biophysics Research Group

    2017-12-15

    The neutron physics characteristic of Pebble Bed Reactor (PBR) allows a better incineration of plutonium (Pu). An optimized design of simple PBR might give a symbiotic solution of providing a safe energy source, effective fuel utilization shown by a higher burnup value, and incineration of Pu stockpiles. This study perform a preliminary neutronic design study of a 200 MWt Once Through Then Out (OTTO) cycle PBR with Pu-based fuel. The safety criteria of the design were represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. In this preliminary phase, the parametric survey is limited to the heavy metal (HM) loading per pebble and the average axial speed of the fuel. An optimum high burnup of 419.7 MWd/kg-HM was achieved in this study. This optimum design uses a HM loading of 2.5 g/pebble with average axial fuel velocity 0.5 cm/day.

  20. Properties of unirradiated fuel element graphites H-451 and SO818. [Bulk density, tensile properties, thermal expansion, thermal conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Engle, G.B.; Johnson, W.R.

    1976-10-08

    Nuclear graphites H-451, lot 440 (Great Lakes Carbon Corporation (GLCC)), and SO818 (Airco Speer Division, Air Reduction Corporation (AS)) are described, and physical, mechanical, and chemical property data are presented for the graphites in the unirradiated state. A summary of the mean values of the property data and of data on TS-1240 and H-451, lot 426, is tabulated. A direct comparison of H-451, lot 426, chosen for Fort St. Vrain (FSV) fuel reload production, TS-1240, and SO818 may be made from the table. (auth)

  1. A safety re-evaluation of the AVR pebble bed reactor operation and its consequences for future HTR concepts

    Energy Technology Data Exchange (ETDEWEB)

    Moormann, R.

    2008-06-15

    The AVR pebble bed reactor (46 MW{sub th}) was operated 1967-88 at coolant outlet temperatures up to 990 C. A principle difference of pebble bed HTRs as AVR to conventional reactors is the continuous movement of fuel element pebbles through the core which complicates thermohydraulic, nuclear and safety estimations. Also because of a lack of other experience AVR operation is still a relevant basis for future pebble bed HTRs and thus requires careful examination. This paper deals mainly with some insufficiently published unresolved safety problems of AVR operation and of pebble bed HTRs but skips the widely known advantageous features of pebble bed HTRs. The AVR primary circuit is heavily contaminated with metallic fission products (Sr-90, Cs-137) which create problems in current dismantling. The amount of this contamination is not exactly known, but the evaluation of fission product deposition experiments indicates that the end of life contamination reached several percent of a single core inventory, which is some orders of magnitude more than precalculated and far more than in large LWRs. A major fraction of this contamination is bound on graphitic dust and thus partly mobile in depressurization accidents, which has to be considered in safety analyses of future reactors. A re-evaluation of the AVR contamination is performed here in order to quantify consequences for future HTRs (400 MW{sub th}). It leads to the conclusion that the AVR contamination was mainly caused by inadmissible high core temperatures, increasing fission product release rates, and not - as presumed in the past - by inadequate fuel quality only. The high AVR core temperatures were detected not earlier than one year before final AVR shut-down, because a pebble bed core cannot yet be equipped with instruments. The maximum core temperatures are still unknown but were more than 200 K higher than calculated. Further, azimuthal temperature differences at the active core margin of up to 200 K were

  2. A safety re-evaluation of the AVR pebble bed reactor operation and its consequences for future HTR concepts

    International Nuclear Information System (INIS)

    Moormann, R.

    2008-06-01

    The AVR pebble bed reactor (46 MW th ) was operated 1967-88 at coolant outlet temperatures up to 990 C. A principle difference of pebble bed HTRs as AVR to conventional reactors is the continuous movement of fuel element pebbles through the core which complicates thermohydraulic, nuclear and safety estimations. Also because of a lack of other experience AVR operation is still a relevant basis for future pebble bed HTRs and thus requires careful examination. This paper deals mainly with some insufficiently published unresolved safety problems of AVR operation and of pebble bed HTRs but skips the widely known advantageous features of pebble bed HTRs. The AVR primary circuit is heavily contaminated with metallic fission products (Sr-90, Cs-137) which create problems in current dismantling. The amount of this contamination is not exactly known, but the evaluation of fission product deposition experiments indicates that the end of life contamination reached several percent of a single core inventory, which is some orders of magnitude more than precalculated and far more than in large LWRs. A major fraction of this contamination is bound on graphitic dust and thus partly mobile in depressurization accidents, which has to be considered in safety analyses of future reactors. A re-evaluation of the AVR contamination is performed here in order to quantify consequences for future HTRs (400 MW th ). It leads to the conclusion that the AVR contamination was mainly caused by inadmissible high core temperatures, increasing fission product release rates, and not - as presumed in the past - by inadequate fuel quality only. The high AVR core temperatures were detected not earlier than one year before final AVR shut-down, because a pebble bed core cannot yet be equipped with instruments. The maximum core temperatures are still unknown but were more than 200 K higher than calculated. Further, azimuthal temperature differences at the active core margin of up to 200 K were observed

  3. Numerical investigations of the fuel cycle for a 10 GW(TH)-OTTO-pebble-bed reactor with regard to high conversion ratio under special consideration of U-236 disconnexion through isotope-separation

    International Nuclear Information System (INIS)

    Werner, H.

    1976-12-01

    A conversion ratio of near 1.0 can be achieved in a pebble-bed reactor using the OTTO (once through then out) loading scheme, having an economic burn-up of the fuel, an economic power density and a moderation ratio, which is considered realistically for the future. The flexibility of the reactor concept and of the fuel element design allows to recycle the fuel during full-power operation. In the present report first the criteria are shown, which are necessary to reach a high conversion ratio. Further it is presented that the conversion ratio increases considerably by closing the fuel cycle in consequence of the building-up of U-233. In this way the fuel inventory and the fuel consumption can considerably be diminished. It is demonstrated that the building-up and the accumulation of U-236 effects an important deterioration of the neutron economy. By taking the reprocessed uranium through an isotope separation (for example: ultra-gas-centrifugation) and by separation of U-236 from the other uranium isotopes it is possible to reduce the fuel consumption considerably. The expenditure and the cost which are necessary for the isotope separation are presented. (orig.) [de

  4. Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

    Energy Technology Data Exchange (ETDEWEB)

    Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

    2009-10-01

    High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next

  5. Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, Yajuan, E-mail: yajuan.zhong@gmail.com [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences, Taiyuan 030001 (China); Zhang, Junpeng [CAS Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences, Taiyuan 030001 (China); Lin, Jun, E-mail: linjun@sinap.ac.cn [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Xu, Liujun [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Guo, Quangui [CAS Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences, Taiyuan 030001 (China)

    2017-07-15

    Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10{sup −6} K{sup −1} (α{sub ∥}) and 6.15 × 10{sup −6} K{sup −1} (α{sub ⊥}) at the temperature range of 25–700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.

  6. Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

    International Nuclear Information System (INIS)

    Zhong, Yajuan; Zhang, Junpeng; Lin, Jun; Xu, Liujun; Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong; Guo, Quangui

    2017-01-01

    Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10 −6 K −1 (α ∥ ) and 6.15 × 10 −6 K −1 (α ⊥ ) at the temperature range of 25–700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.

  7. Localization of the hot spots in a pebble bed reactor

    International Nuclear Information System (INIS)

    Chen, Leisheng; Lee, Wooram; Lee, Jaeyoung

    2016-01-01

    The pebble bed reactor (PBR) is a candidate reactor type for the very high temperature reactor (VHTR), which is one of the Generation-IV reactor types. The HTGR design concept exhibits excellent safety features due to the low power density and the large amount of graphite present in the core which gives a large thermal inertia in an accident such as loss of coolant. The conclusions are made and may contribute to a better design of a PBR core and a closer inspection of the local hot spots to avoid destruction of pebbles from happening. Thermal field of a PBR core is investigated in this study. Specifically, experiments on measuring the pebbles' surface temperature are performed. It is found that the upper pebble has an overall higher temperature profile than the other pebbles and the stagnation zone under does not increase its surface's temperature. In addition, the temperature profile of the side pebble shows a concave form and it keeps decreasing from the contact point to the vertex in the lower pebble. Lastly, the maximum temperature difference among these points is 5.83 deg. C. These findings above are validated by CFX simulations under two different turbulence models (k-e, SST) and two contact areas (diameter of 6mm and 3.5mm). By contrasting the temperature variation trends of all simulation cases, it is concluded that SST turbulence model with 20% intensity shows a better agreement with the experiment result, nevertheless, slightly deviation is also found in terms of total temperature difference and the peak appears in position 17-19 in experiments

  8. Localization of the hot spots in a pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Leisheng; Lee, Wooram; Lee, Jaeyoung [Handong Global University, Pohang (Korea, Republic of)

    2016-05-15

    The pebble bed reactor (PBR) is a candidate reactor type for the very high temperature reactor (VHTR), which is one of the Generation-IV reactor types. The HTGR design concept exhibits excellent safety features due to the low power density and the large amount of graphite present in the core which gives a large thermal inertia in an accident such as loss of coolant. The conclusions are made and may contribute to a better design of a PBR core and a closer inspection of the local hot spots to avoid destruction of pebbles from happening. Thermal field of a PBR core is investigated in this study. Specifically, experiments on measuring the pebbles' surface temperature are performed. It is found that the upper pebble has an overall higher temperature profile than the other pebbles and the stagnation zone under does not increase its surface's temperature. In addition, the temperature profile of the side pebble shows a concave form and it keeps decreasing from the contact point to the vertex in the lower pebble. Lastly, the maximum temperature difference among these points is 5.83 deg. C. These findings above are validated by CFX simulations under two different turbulence models (k-e, SST) and two contact areas (diameter of 6mm and 3.5mm). By contrasting the temperature variation trends of all simulation cases, it is concluded that SST turbulence model with 20% intensity shows a better agreement with the experiment result, nevertheless, slightly deviation is also found in terms of total temperature difference and the peak appears in position 17-19 in experiments.

  9. Parametric study for high conversion pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Teuchert, E.; Ruetten, H. J.

    1975-06-15

    Tables are presented of fuel cycle costs, conversion ratios and accompanying variations in fuel element designs for a 3,00 MWth high conversion pebble bed reactor with initial high enriched uranium/thorium cycle and subsequent recycling of U-233, Pu-239 and Pu-241.

  10. An analytical evaluation for spatial-dependent intra-pebble Dancoff factor and escape probability

    International Nuclear Information System (INIS)

    Kim, Songhyun; Kim, Hong-Chul; Kim, Jong Kyung; Kim, Soon Young; Noh, Jae Man

    2009-01-01

    The analytical evaluation of spatial-dependent intra-pebble Dancoff factors and their escape probabilities is pursued by the model developed in this study. Intra-pebble Dancoff factors and their escape probabilities are calculated as a function of fuel kernel radius, number of fuel kernels, and fuel region radius. The method in this study can be easily utilized to analyze the tendency of spatial-dependent intra-pebble Dancoff factor and spatial-dependent fuel region escape probability for the various geometries because it is faster than the MCNP method as well as good accuracy. (author)

  11. Review of PSI studies on reactor physics and thermal fluid dynamics of pebble bed reactors

    International Nuclear Information System (INIS)

    Prasser, Horst-Michael

    2014-01-01

    density ratios are expected in air ingress scenarios at an HTR. As a spin-off from aerosol studies for severe accidents, theoretical studies were started on graphite dust in pebble bed reactors. Wear and tear of the fuel spheres can produce significant quantities of graphite dust. Simulations of the pebble flow in a random package have been carried out in a generic full-size reactor geometry (440’000 pebbles) using the discrete-element method (DEM). The simulations provide the residence time distribution of the pebbles and the spatial distribution of wear. The model is ready for an implementation of a dust production term. In parallel, preliminary results on dust deposition were obtained for a particle laden flow around a single sphere and a linear arrangement of spheres. For this purpose a RANS turbulence model was coupled with a continuous random walk model for the integration of the particle trajectories in Lagrangian coordinates. Based on the outlined expertise, PSI plans to intensify the work on HTR in the future. (author)

  12. Applying burnable poison particles to reduce the reactivity swing in high temperature reactors with batch-wise fuel loading

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Dam, H. van; Hagen, T.H.J.J. van der

    2003-01-01

    Burnup calculations have been performed on a standard HTR fuel pebble with a radius of 3 cm containing 9 g of 8% enriched uranium and burnable poison particles (BPP) made of B 4 C highly enriched in 10 B. The radius of the BPP and the number of particles per fuel pebble have been varied to find the flattest reactivity-to-time curve. It was found that for a k∞ of 1.1, a reactivity swing as low as 2% can be obtained when each fuel pebble contains about 1070 BPP with a radius of 75 μm. For coated BPP that consist of a graphite kernel with a radius of 300 μm covered with a B 4 C burnable poison layer, a similar value for the reactivity swing can be obtained. Cylindrical particles seem to perform worse. In general, the modification of the geometry of BPP is an effective means to tailor the reactivity curve of HTRs

  13. 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.

  14. 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

  15. Thermal safety analysis for pebble bed blanket fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Wei Renjie

    1998-01-01

    Pebble bed blanket hybrid reactor may have more advantages than slab element blanket hybrid reactor in nuclear fuel production and nuclear safety. The thermo-hydraulic calculations of the blanket in the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor developed in China are carried out using the Code THERMIX and auxiliary code. In the calculations different fuel pebble material and steady state, depressurization and total loss of flow accident conditions are included. The results demonstrate that the conceptual design of the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor with dump tank is feasible and safe enough only if the suitable fuel pebble material is selected and the suitable control system and protection system are established. Some recommendations for due conceptual design are also presented

  16. A Pebble Bed Reactor cross section methodology

    International Nuclear Information System (INIS)

    Hudson, Nathanael H.; Ougouag, Abderrafi M.; Rahnema, Farzad; Gougar, Hans

    2009-01-01

    A method is presented for the evaluation of microscopic cross sections for the Pebble Bed Reactor (PBR) neutron diffusion computational models during convergence to an equilibrium (asymptotic) fuel cycle. This method considers the isotopics within a core spectral zone and the leakages from such a zone as they arise during reactor operation. The randomness of the spatial distribution of fuel grains within the fuel pebbles and that of the fuel and moderator pebbles within the core, the double heterogeneity of the fuel, and the indeterminate burnup of the spectral zones all pose a unique challenge for the computation of the local microscopic cross sections. As prior knowledge of the equilibrium composition and leakage is not available, it is necessary to repeatedly re-compute the group constants with updated zone information. A method is presented to account for local spectral zone composition and leakage effects without resorting to frequent spectrum code calls. Fine group data are pre-computed for a range of isotopic states. Microscopic cross sections and zone nuclide number densities are used to construct fine group macroscopic cross sections, which, together with fission spectra, flux modulation factors, and zone buckling, are used in the solution of the slowing down balance to generate a new or updated spectrum. The microscopic cross-sections are then re-collapsed with the new spectrum for the local spectral zone. This technique is named the Spectral History Correction (SHC) method. It is found that this method accurately recalculates local broad group microscopic cross sections. Significant improvement in the core eigenvalue, flux, and power peaking factor is observed when the local cross sections are corrected for the effects of the spectral zone composition and leakage in two-dimensional PBR test problems.

  17. Pebbles, Cobbles, and Sockets

    Science.gov (United States)

    1997-01-01

    This Rover image of 'Shark' (upper left center), 'Half Dome' (upper right), and a small rock (right foreground) reveal textures and structures not visible in lander camera images. These rocks are interpreted as conglomerates because their surfaces have rounded protrusions up to several centimeters in size. It is suggested that the protrusions are pebbles and granules.Mars Pathfinder is the second in NASA's Discovery program of low-cost spacecraft with highly focused science goals. The Jet Propulsion Laboratory, Pasadena, CA, developed and manages the Mars Pathfinder mission for NASA's Office of Space Science, Washington, D.C. JPL is a division of the California Institute of Technology (Caltech).

  18. Calculation of the packing fraction in a pebble-bed ADS and redesigning of the Transmutation Advanced Device for Sustainable Energy Applications (TADSEA)

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, L., E-mail: maiden@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), Av. Salvador Allende y Luaces, Ciudad de la Habana, 10400 (Cuba); Perez, J., E-mail: jcurbelo@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), Av. Salvador Allende y Luaces, Ciudad de la Habana, 10400 (Cuba); Garcia, C., E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), Av. Salvador Allende y Luaces, Ciudad de la Habana, 10400 (Cuba); Escriva, A., E-mail: aescriva@iqn.upv.es [Instituto de Ingenieria Energetica (IIE), Universidad Politecnica de Valencia (UPV), Camino de Vera s/n, 46022 Valencia (Spain); Rosales, J., E-mail: jrosales@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), Av. Salvador Allende y Luaces, Ciudad de la Habana, 10400 (Cuba); Abanades, A., E-mail: abanades@etsii.upm.es [Escuela Superior de Ingenieros Industriales (ETSII), Universidad Politecnica de Madrid (UPM), J. Gutierrez Abascal, 2, 28006 Madrid (Spain)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer We based our study on an ADS for TRU transmutation and high temperature production. Black-Right-Pointing-Pointer We calculated the number of pebbles that fit in a cylindrical ADS core. Black-Right-Pointing-Pointer In both ADS design options studied, the mass of Pu isotopes reduces considerably. Black-Right-Pointing-Pointer The system can reach coolant outlet temperatures high enough for hydrogen production. Black-Right-Pointing-Pointer The maximum temperature values obtained in the ADS are not dangerous for TRISO fuel. - Abstract: One of the main problems that should be addressed in the use of nuclear fuels for heat and electricity production is the management of nuclear waste from conventional nuclear power plants and its inventory minimization. Fast reactors and Accelerator Driven Systems (ADSs) are the main options for reducing the long-lived radioactive waste inventory. In previous studies, the conceptual design of a Transmutation Advanced Device for Sustainable Energy Applications (TADSEA) has been made. The TADSEA is a pebble-bed ADS cooled by helium and moderated by graphite; it uses as fuel small amounts of transuranic elements in the form of TRISO particles, confined in 3 cm radius graphite pebbles. It has been conceived for Plutonium (Pu) and Minor Actinides (MA) transmutation and for achieving very high helium temperatures at the core's outlet to match the thermal requirements for hydrogen production by high temperature electrolysis (HTE) or by the iodine-sulfur (I-S) thermo-chemical cycle. In this paper, a geometrical method for calculating the real number of pebbles that fit in a cylindrical ADS core, according to its size and pebble configuration, is described. Based on its results, the packing fraction influence on the TADSEA's main work parameters is studied, and the redesign of the previous configuration is done in order to maintain the exit thermal power established in the preliminary design

  19. Effect of NaX zeolite-modified graphite felts on hexavalent chromium removal in biocathode microbial fuel cells.

    Science.gov (United States)

    Wu, Xiayuan; Tong, Fei; Yong, Xiaoyu; Zhou, Jun; Zhang, Lixiong; Jia, Honghua; Wei, Ping

    2016-05-05

    Two kinds of NaX zeolite-modified graphite felts were used as biocathode electrodes in hexavalent chromium (Cr(VI))-reducing microbial fuel cells (MFCs). The one was fabricated through direct modification, and the other one processed by HNO3 pretreatment of graphite felt before modification. The results showed that two NaX zeolite-modified graphite felts are excellent bio-electrode materials for MFCs, and that a large NaX loading mass, obtained by HNO3 pretreatment (the HNO3-NaX electrode), leads to a superior performance. The HNO3-NaX electrode significantly improved the electricity generation and Cr(VI) removal of the MFC. The maximum Cr(VI) removal rate increased to 10.39±0.28 mg/L h, which was 8.2 times higher than that of the unmodified control. The improvement was ascribed to the strong affinity that NaX zeolite particles, present in large number on the graphite felt, have for microorganisms and Cr(VI) ions. Copyright © 2016 Elsevier B.V. All rights reserved.

  20. Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled High Temperature Reactor - 15171

    International Nuclear Information System (INIS)

    Zhu, G.; Zou, Y.; Xu, Hongjie

    2015-01-01

    Sustainability of thorium fuel in a pebble-bed fluoride salt-cooled high temperature reactor (PB-FHR) is investigated to find the feasible region of high discharge burnup and negative FLiBe (2LiF-BeF 2 ) salt temperature reactivity coefficient (TRC). Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tri-structural-isotropic (TRISO) coated particle system for increasing heavy metal loading and decreasing excessive moderation. In order to analyze the neutronic characteristics, an equilibrium calculation method of thorium fuel self-sustainability is developed. We have compared 2 refueling schemes (mixing flow pattern and directional flow pattern) and 2 kinds of reflector materials (SiC and graphite). This method has found that the feasible regions of breeding and negative FLiBe TRC is between 20 vol% and 62 vol% heavy metal loading in the fuel. A discharge burnup could be achieved up to about 200 MWd/kgHM. The case with directional flow pattern and SiC reflector showed superior burnup characteristics but the worst radial power peak factor, while the case with mixing flow pattern and SiC reflector, which was the best tradeoff between discharge burnup and radial power peak factor, could provide burnup of 140 MWd/kgHM and about 1.4 radial power peak factor with 50 vol% dispersion fuel. In addition, FLiBe salt displays good neutron properties as a coolant of quasi-fast reactors due to the strong 9 Be(n,2n) reaction and low neutron absorption of 6 Li (even at 1000 ppm) in fast spectrum. Preliminary thermal hydraulic calculation shows good safety margins. The greatest challenge of this reactor may be the very long irradiation time of the pebble fuel. (authors)

  1. Brazing graphite to graphite

    International Nuclear Information System (INIS)

    Peterson, G.R.

    1976-01-01

    Graphite is joined to graphite by employing both fine molybdenum powder as the brazing material and an annealing step that together produce a virtually metal-free joint exhibiting properties similar to those found in the parent graphite. Molybdenum powder is placed between the faying surfaces of two graphite parts and melted to form molybdenum carbide. The joint area is thereafter subjected to an annealing operation which diffuses the carbide away from the joint and into the graphite parts. Graphite dissolved by the dispersed molybdenum carbide precipitates into the joint area, replacing the molybdenum carbide to provide a joint of graphite

  2. Organic-resistant screen-printed graphitic electrodes: Application to on-site monitoring of liquid fuels

    International Nuclear Information System (INIS)

    Almeida, Eduardo S.; Silva, Luiz A.J.; Sousa, Raquel M.F.; Richter, Eduardo M.; Foster, Christopher W.; Banks, Craig E.; Munoz, Rodrigo A.A.

    2016-01-01

    This work presents the potential application of organic-resistant screen-printed graphitic electrodes (SPGEs) for fuel analysis. The required analysis of the antioxidant 2,6-di-tert-butylphenol (2,6-DTBP) in biodiesel and jet fuel is demonstrated as a proof-of-concept. The screen-printing of graphite, Ag/AgCl and insulator inks on a polyester substrate (250 μm thickness) resulted in SPGEs highly compatible with liquid fuels. SPGEs were placed on a batch-injection analysis (BIA) cell, which was filled with a hydroethanolic solution containing 99% v/v ethanol and 0.1 mol L −1 HClO 4 (electrolyte). An electronic micropipette was connected to the cell to perform injections (100 μL) of sample or standard solutions. Over 200 injections can be injected continuously without replacing electrolyte and SPGE strip. Amperometric detection (+1.1 V vs. Ag/AgCl) of 2,6-DTBP provided fast (around 8 s) and precise (RSD = 0.7%, n = 12) determinations using an external calibration curve. The method was applied for the analysis of biodiesel and aviation jet fuel samples and comparable results with liquid and gas chromatographic analyses, typically required for biodiesel and jet fuel samples, were obtained. Hence, these SPGE strips are completely compatible with organic samples and their combination with the BIA cell shows great promise for routine and portable analysis of fuels and other organic liquid samples without requiring sophisticated sample treatments. - Highlights: • Organic-resistant screen-printed graphitic electrodes (SPGE) for (bio)fuels. • Screen-printing of conductive and insulator inks on thin polyester substrate. • Continuous detection of antioxidants in electrolyte with 99% v/v ethanol. • SPGE coupled with batch-injection analysis allows over 200 injections (100 μL). • Similar results to GC and HPLC analyses of biodiesel and aviation jet fuels.

  3. Organic-resistant screen-printed graphitic electrodes: Application to on-site monitoring of liquid fuels

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, Eduardo S.; Silva, Luiz A.J.; Sousa, Raquel M.F.; Richter, Eduardo M. [Universidade Federal de Uberlândia, Universidade Federal de Uberlândia, Av. João Naves de Ávila, 2121, Uberlândia, MG, 38408100 (Brazil); Foster, Christopher W.; Banks, Craig E. [Manchester Metropolitan University, Faculty of Science and the Environment, School of Science and the Environment, Division of Chemistry and Environmental Science, Manchester, M1 5GD, England (United Kingdom); Munoz, Rodrigo A.A., E-mail: raamunoz@iqufu.ufu.br [Universidade Federal de Uberlândia, Universidade Federal de Uberlândia, Av. João Naves de Ávila, 2121, Uberlândia, MG, 38408100 (Brazil)

    2016-08-31

    This work presents the potential application of organic-resistant screen-printed graphitic electrodes (SPGEs) for fuel analysis. The required analysis of the antioxidant 2,6-di-tert-butylphenol (2,6-DTBP) in biodiesel and jet fuel is demonstrated as a proof-of-concept. The screen-printing of graphite, Ag/AgCl and insulator inks on a polyester substrate (250 μm thickness) resulted in SPGEs highly compatible with liquid fuels. SPGEs were placed on a batch-injection analysis (BIA) cell, which was filled with a hydroethanolic solution containing 99% v/v ethanol and 0.1 mol L{sup −1} HClO{sub 4} (electrolyte). An electronic micropipette was connected to the cell to perform injections (100 μL) of sample or standard solutions. Over 200 injections can be injected continuously without replacing electrolyte and SPGE strip. Amperometric detection (+1.1 V vs. Ag/AgCl) of 2,6-DTBP provided fast (around 8 s) and precise (RSD = 0.7%, n = 12) determinations using an external calibration curve. The method was applied for the analysis of biodiesel and aviation jet fuel samples and comparable results with liquid and gas chromatographic analyses, typically required for biodiesel and jet fuel samples, were obtained. Hence, these SPGE strips are completely compatible with organic samples and their combination with the BIA cell shows great promise for routine and portable analysis of fuels and other organic liquid samples without requiring sophisticated sample treatments. - Highlights: • Organic-resistant screen-printed graphitic electrodes (SPGE) for (bio)fuels. • Screen-printing of conductive and insulator inks on thin polyester substrate. • Continuous detection of antioxidants in electrolyte with 99% v/v ethanol. • SPGE coupled with batch-injection analysis allows over 200 injections (100 μL). • Similar results to GC and HPLC analyses of biodiesel and aviation jet fuels.

  4. Neutronic modeling of pebble bed reactors in APOLLO2

    International Nuclear Information System (INIS)

    Grimod, M.

    2010-01-01

    In this thesis we develop a new iterative homogenization technique for pebble bed reactors, based on a 'macro-stochastic' transport approximation in the collision probability method. A model has been developed to deal with the stochastic distribution of pebbles with different burnup in the core, considering spectral differences in homogenization and depletion calculations. This is generally not done in the codes presently used for pebble bed analyses, where a pebble with average isotopic composition is considered to perform the cell calculation. Also an iterative core calculation scheme has been set up, where the low-order RZ S N full-core calculation computes the entering currents in the spectrum zones subdividing the core. These currents, together with the core k eff , are then used as surface source in the fine-group heterogeneous calculation of the multi-pebble geometries. The developed method has been verified using reference Monte Carlo simulations of a simplified PBMR- 400 model. The pebbles in this model are individually positioned and have different randomly assigned burnup values. The APOLLO2 developed method matches the reference core k eff within ± 100 pcm, with relative differences on the production shape factors within ± 4%, and maximum discrepancy of 3% at the hotspot. Moreover, the first criticality experiment of the HTR-10 reactor was used to perform a first validation of the developed model. The computed critical number of pebbles to be loaded in the core is very close to the experimental value of 16890, only 77 pebbles less. A method to calculate the equilibrium reactor state was also developed and applied to analyze the simplified PBMR-400 model loaded with different fuel types (UO 2 , Pu, Pu + MA). The potential of the APOLLO2 method to compute different fluxes for the different pebble types of a multi-pebble geometry was used to evaluate the bias committed by the average composition pebble approximation. Thanks to a 'compensation of error

  5. Organic-resistant screen-printed graphitic electrodes: Application to on-site monitoring of liquid fuels.

    Science.gov (United States)

    Almeida, Eduardo S; Silva, Luiz A J; Sousa, Raquel M F; Richter, Eduardo M; Foster, Christopher W; Banks, Craig E; Munoz, Rodrigo A A

    2016-08-31

    This work presents the potential application of organic-resistant screen-printed graphitic electrodes (SPGEs) for fuel analysis. The required analysis of the antioxidant 2,6-di-tert-butylphenol (2,6-DTBP) in biodiesel and jet fuel is demonstrated as a proof-of-concept. The screen-printing of graphite, Ag/AgCl and insulator inks on a polyester substrate (250 μm thickness) resulted in SPGEs highly compatible with liquid fuels. SPGEs were placed on a batch-injection analysis (BIA) cell, which was filled with a hydroethanolic solution containing 99% v/v ethanol and 0.1 mol L(-1) HClO4 (electrolyte). An electronic micropipette was connected to the cell to perform injections (100 μL) of sample or standard solutions. Over 200 injections can be injected continuously without replacing electrolyte and SPGE strip. Amperometric detection (+1.1 V vs. Ag/AgCl) of 2,6-DTBP provided fast (around 8 s) and precise (RSD = 0.7%, n = 12) determinations using an external calibration curve. The method was applied for the analysis of biodiesel and aviation jet fuel samples and comparable results with liquid and gas chromatographic analyses, typically required for biodiesel and jet fuel samples, were obtained. Hence, these SPGE strips are completely compatible with organic samples and their combination with the BIA cell shows great promise for routine and portable analysis of fuels and other organic liquid samples without requiring sophisticated sample treatments. Copyright © 2016 Elsevier B.V. All rights reserved.

  6. Special graphites; Graphites speciaux

    Energy Technology Data Exchange (ETDEWEB)

    Leveque, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [French] Ameliorer les proprietes du graphite nucleaire pour empilements et ouvrir de nouveaux domaines d'application au graphite constituent une part importante de l'effort entrepris en commun par le Commissariat a l'Energie Atomique (CEA) et la compagnie PECHINEY. Des procedes nouveaux de fabrication de carbones et graphites speciaux ont ete mis au point: graphite forge, pyrocarbone, graphite de haute densite, agglomeration de poudres de graphite par craquage de gaz naturel, graphites impermeables. Les proprietes physiques de ces produits ainsi que leur reaction avec differents gaz oxydants sont decrites. Les premiers resultats d'irradiation sont aussi donnes. (auteurs)

  7. Computational and experimental prediction of dust production in pebble bed reactors, Part II

    Energy Technology Data Exchange (ETDEWEB)

    Mie Hiruta; Gannon Johnson; Maziar Rostamian; Gabriel P. Potirniche; Abderrafi M. Ougouag; Massimo Bertino; Louis Franzel; Akira Tokuhiro

    2013-10-01

    This paper is the continuation of Part I, which describes the high temperature and high pressure helium environment wear tests of graphite–graphite in frictional contact. In the present work, it has been attempted to simulate a Pebble Bed Reactor core environment as compared to Part I. The experimental apparatus, which is a custom-designed tribometer, is capable of performing wear tests at PBR relevant higher temperatures and pressures under a helium environment. This environment facilitates prediction of wear mass loss of graphite as dust particulates from the pebble bed. The experimental results of high temperature helium environment are used to anticipate the amount of wear mass produced in a pebble bed nuclear reactor.

  8. Graphite anode surface modification with controlled reduction of specific aryl diazonium salts for improved microbial fuel cells power output.

    Science.gov (United States)

    Picot, Matthieu; Lapinsonnière, Laure; Rothballer, Michael; Barrière, Frédéric

    2011-10-15

    Graphite electrodes were modified with reduction of aryl diazonium salts and implemented as anodes in microbial fuel cells. First, reduction of 4-aminophenyl diazonium is considered using increased coulombic charge density from 16.5 to 200 mC/cm(2). This procedure introduced aryl amine functionalities at the surface which are neutral at neutral pH. These electrodes were implemented as anodes in "H" type microbial fuel cells inoculated with waste water, acetate as the substrate and using ferricyanide reduction at the cathode and a 1000 Ω external resistance. When the microbial anode had developed, the performances of the microbial fuel cells were measured under acetate saturation conditions and compared with those of control microbial fuel cells having an unmodified graphite anode. We found that the maximum power density of microbial fuel cell first increased as a function of the extent of modification, reaching an optimum after which it decreased for higher degree of surface modification, becoming even less performing than the control microbial fuel cell. Then, the effect of the introduction of charged groups at the surface was investigated at a low degree of surface modification. It was found that negatively charged groups at the surface (carboxylate) decreased microbial fuel cell power output while the introduction of positively charged groups doubled the power output. Scanning electron microscopy revealed that the microbial anode modified with positively charged groups was covered by a dense and homogeneous biofilm. Fluorescence in situ hybridization analyses showed that this biofilm consisted to a large extent of bacteria from the known electroactive Geobacter genus. In summary, the extent of modification of the anode was found to be critical for the microbial fuel cell performance. The nature of the chemical group introduced at the electrode surface was also found to significantly affect the performance of the microbial fuel cells. The method used for

  9. Graphite selection for the PBMR reflector

    International Nuclear Information System (INIS)

    Marsden, B.J.; Preston, S.D.

    2000-01-01

    A high temperature, direct cycle gas turbine, graphite moderated, helium cooled, pebble-bed reactor (PBMR) is being designed and constructed in South Africa. One of the major components in the PBMR is the graphite reflector, which must be designed to last thirty-five full power years. Fast neutron irradiation changes the dimensions and material properties of reactor graphite, thus for design purposes a suitable graphite database is required. Data on the effect of irradiation on nuclear graphites has been gathered for many years, at considerable financial cost, but unfortunately these graphites are no longer available due to rationalization of the graphite industry and loss of key graphite coke supplies. However, it is possible, using un-irradiated graphite materials properties and knowledge of the particular graphite microstructure, to determine the probable irradiation behaviour. Three types of nuclear graphites are currently being considered for the PBMR reflector: an isostatically moulded, fine grained, high strength graphite and two extruded medium grained graphites of moderately high strength. Although there is some irradiation data available for these graphites, the data does not cover the temperature and dose range required for the PBMR. The available graphites have been examined to determine their microstructure and some of the key material properties are presented. (authors)

  10. Spontaneous modification of graphite anode by anthraquinone-2-sulfonic acid for microbial fuel cells.

    Science.gov (United States)

    Tang, Xinhua; Li, Haoran; Du, Zhuwei; Ng, How Yong

    2014-07-01

    In this study, anthraquinone-2-sulfonic acid (AQS), an electron transfer mediator, was immobilized onto graphite felt surface via spontaneous reduction of the in situ generated AQS diazonium cations. Cyclic voltammetry (CV) and energy dispersive spectrometry (EDS) characterizations of AQS modified graphite demonstrated that AQS was covalently grafted onto the graphite surface. The modified graphite, with a surface AQS concentration of 5.37 ± 1.15 × 10(-9)mol/cm(2), exhibited good electrochemical activity and high stability. The midpoint potential of the modified graphite was about -0.248 V (vs. normal hydrogen electrode, NHE), indicating that electrons could be easily transferred from NADH in bacteria to the electrode. AQS modified anode in MFCs increased the maximum power density from 967 ± 33 mW/m(2) to 1872 ± 42 mW/m(2). These results demonstrated that covalently modified AQS functioned as an electron transfer mediator to facilitate extracellular electron transfer from bacteria to electrode and significantly enhanced the power production in MFCs. Copyright © 2014 Elsevier Ltd. All rights reserved.

  11. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    International Nuclear Information System (INIS)

    Mohanty, Subhasish; Majumdar, Saurindranath; Srinivasan, Makuteswara

    2013-01-01

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite

  12. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Majumdar, Saurindranath [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission, Washington, DC 20555 (United States)

    2013-07-15

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite.

  13. Neutron physical investigations on the shutdown effect of small boronated absorbing spheres for pebble-bed high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Sgouridis, S.; Schurrer, F.; Muller, H.; Ninaus, W.; Oswald, K.; Neef, R.D.; Schaal, H.

    1987-01-01

    An emergency shutdown system for high-temperature gas-cooled pebble-bed reactors is proposed in addition to the common absorber rod shutdown system. This system is based on the strongly absorbing effect of small boronated graphite spheres (called KLAK), which trickle in case of emergency by gravity from the top reflector into the reactor core. The inner reflector of the Siemens-Argonaut reactor was substituted by an assembly of spherical Arbeitsgemeinschaft Versuchsreaktor fuel elements, and the shutdown effect was examined by installing well-defined KLAK nests inside this assembly. The purpose was to develop and prove a calculational procedure for determining criticality values for assemblies of large fuel spheres and small absorbing spheres

  14. AGC-2 Graphite Preirradiation Data Package

    Energy Technology Data Exchange (ETDEWEB)

    David Swank; Joseph Lord; David Rohrbaugh; William Windes

    2012-10-01

    The NGNP Graphite R&D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterized prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.

  15. Fuel management of HTR-10

    International Nuclear Information System (INIS)

    Wu Zongxin; Jing Xingqing

    2001-01-01

    The 10 MW high temperature cooled reactor (HTR-10) built in Tsinghua University is a pebble bed type of HTGR. The continuous recharge and multiple-pass of spherical fuel elements are used for fuel management. The initiative stage of core is composed of the mix of spherical fuel elements and graphite elements. The equilibrium stage of core is composed of identical spherical fuel elements. The fuel management during the transition from the initiative stage to the equilibrium stage is a key issue for HTR-10 physical design. A fuel management strategy is proposed based on self-adjustment of core reactivity. The neutron physical code is used to simulate the process of fuel management. The results show that the graphite elements, the recharging fuel elements below the burn-up allowance, and the discharging fuel elements over the burn-up allowance could be identified by burn-up measurement. The maximum of burn-up fuel elements could be controlled below the burn-up limit

  16. The Emerging Paradigm of Pebble Accretion

    NARCIS (Netherlands)

    Ormel, C.W.; Pessah, M.; Gressel, O.

    2017-01-01

    Pebble accretion is the mechanism in which small particles ("pebbles") accrete onto big bodies big (planetesimals or planetary embryos) in gas-rich environments. In pebble accretion accretion , accretion occurs by settling and depends only on the mass of the gravitating body gravitating , not its

  17. Flow distribution of pebble bed high temperature gas cooled reactors using large eddy simulation

    International Nuclear Information System (INIS)

    Gokhan Yesilyurt; Hassan, Y.A.

    2003-01-01

    A High Temperature Gas-cooled Reactor (HTGR) is one of the renewed reactor designs to play a role in nuclear power generation. This reactor design concepts is currently under consideration and development worldwide. Since the HTGR concept offers inherent safety, has a very flexible fuel cycle with capability to achieve high burnup levels, and provides good thermal efficiency of power plant, it can be considered for further development and improvement as a reactor concept of generation IV. The combination of coated particle fuel, inert helium gas as coolant and graphite moderated reactor makes it possible to operate at high temperature yielding a high efficiency. In this study the simulation of turbulent transport for the gas through the gaps of the spherical fuel elements (fuel pebbles) will be performed. This will help in understanding the highly three-dimensional, complex flow phenomena in pebble bed caused by flow curvature. Under these conditions, heat transfer in both laminar and turbulent flows varies noticeably around curved surfaces. Curved flows would be present in the presence of contiguous curved surfaces. In the case of a laminar flow and of an appreciable effect of thermogravitional forces, the Nusselt (Nu) number depends significantly on the curvature shape of the surface. It changes with order of 10 times. The flow passages through the gap between the fuel balls have concave and convex configurations. Here the action of the centrifugal forces manifests itself differently on convex and concave parts of the flow path (suppression or stimulation of turbulence). The flow of this type has distinctive features. In such flow there is a pressure gradient, which strongly affects the boundary layer behavior. The transition from a laminar to turbulent flow around this curved flow occurs at deferent Reynolds (Re) numbers. Consequently, noncircular curved flows as in the pebble-bed situation, in detailed local sense, is interesting to be investigated. To the

  18. Effects of Boron and Graphite Uncertainty in Fuel for TREAT Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Vaughn, Kyle; Mausolff, Zander; Gonzalez, Esteban; DeHart, Mark; Goluoglu, Sedat

    2017-03-01

    Advanced modeling techniques and current computational capacity make full core TREAT simulations possible, with the goal of such simulations to understand the pre-test core and minimize the number of required calibrations. But, in order to simulate TREAT with a high degree of precision the reactor materials and geometry must also be modeled with a high degree of precision. This paper examines how uncertainty in the reported values of boron and graphite have an effect on simulations of TREAT.

  19. Uncertainty and Sensitivity Analyses of a Pebble Bed HTGR Loss of Cooling Event

    Directory of Open Access Journals (Sweden)

    Gerhard Strydom

    2013-01-01

    Full Text Available The Very High Temperature Reactor Methods Development group at the Idaho National Laboratory identified the need for a defensible and systematic uncertainty and sensitivity approach in 2009. This paper summarizes the results of an uncertainty and sensitivity quantification investigation performed with the SUSA code, utilizing the International Atomic Energy Agency CRP 5 Pebble Bed Modular Reactor benchmark and the INL code suite PEBBED-THERMIX. Eight model input parameters were selected for inclusion in this study, and after the input parameters variations and probability density functions were specified, a total of 800 steady state and depressurized loss of forced cooling (DLOFC transient PEBBED-THERMIX calculations were performed. The six data sets were statistically analyzed to determine the 5% and 95% DLOFC peak fuel temperature tolerance intervals with 95% confidence levels. It was found that the uncertainties in the decay heat and graphite thermal conductivities were the most significant contributors to the propagated DLOFC peak fuel temperature uncertainty. No significant differences were observed between the results of Simple Random Sampling (SRS or Latin Hypercube Sampling (LHS data sets, and use of uniform or normal input parameter distributions also did not lead to any significant differences between these data sets.

  20. Study on Characteristic of Temperature Coefficient of Reactivity for Plutonium Core of Pebbled Bed Reactor

    Science.gov (United States)

    Zuhair; Suwoto; Setiadipura, T.; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    As a part of the solution searching for possibility to control the plutonium, a current effort is focused on mechanisms to maximize consumption of plutonium. Plutonium core solution is a unique case in the high temperature reactor which is intended to reduce the accumulation of plutonium. However, the safety performance of the plutonium core which tends to produce a positive temperature coefficient of reactivity should be examined. The pebble bed inherent safety features which are characterized by a negative temperature coefficient of reactivity must be maintained under any circumstances. The purpose of this study is to investigate the characteristic of temperature coefficient of reactivity for plutonium core of pebble bed reactor. A series of calculations with plutonium loading varied from 0.5 g to 1.5 g per fuel pebble were performed by the MCNPX code and ENDF/B-VII library. The calculation results show that the k eff curve of 0.5 g Pu/pebble declines sharply with the increase in fuel burnup while the greater Pu loading per pebble yields k eff curve declines slighter. The fuel with high Pu content per pebble may reach long burnup cycle. From the temperature coefficient point of view, it is concluded that the reactor containing 0.5 g-1.25 g Pu/pebble at high burnup has less favorable safety features if it is operated at high temperature. The use of fuel with Pu content of 1.5 g/pebble at high burnup should be considered carefully from core safety aspect because it could affect transient behavior into a fatal accident situation.

  1. Effect of graphite addition into mill scale waste as a potential bipolar plates material of proton exchange membrane fuel cells

    Science.gov (United States)

    Khaerudini, D. S.; Prakoso, G. B.; Insiyanda, D. R.; Widodo, H.; Destyorini, F.; Indayaningsih, N.

    2018-03-01

    Bipolar plates (BPP) is a vital component of proton exchange membrane fuel cells (PEMFC), which supplies fuel and oxidant to reactive sites, remove reaction products, collects produced current and provide mechanical support for the cells in the stack. This work concerns the utilization of mill scale, a by-product of iron and steel formed during the hot rolling of steel, as a potential material for use as BPP in PEMFC. On the other hand, mill scale is considered a very rich in iron source having characteristic required such as for current collector in BPP and would significantly contribute to lower the overall cost of PEMFC based fuel cell systems. In this study, the iron reach source of mill scale powder, after sieving of 150 mesh, was mechanically alloyed with the carbon source containing 5, 10, and 15 wt.% graphite using a shaker mill for 3 h. The mixed powders were then pressed at 300 MPa and sintered at 900 °C for 1 h under inert gas atmosphere. The structural changes of powder particles during mechanical alloying and after sintering were studied by X-ray diffractometry, optical microscopy, scanning electron microscopy, and microhardness measurement. The details of the presence of iron, carbon, and iron carbide (Fe-C) as the products of reactions as well as sufficient mechanical strength of the sintered materials were presented in this report.

  2. Effect of anode polarization on biofilm formation and electron transfer in Shewanella oneidensis/graphite felt microbial fuel cells.

    Science.gov (United States)

    Pinto, David; Coradin, Thibaud; Laberty-Robert, Christel

    2018-04-01

    In microbial fuel cells, electricity generation is assumed by bacterial degradation of low-grade organics generating electrons that are transferred to an electrode. The nature and efficiency of the electron transfer from the bacteria to the electrodes are determined by several chemical, physical and biological parameters. Specifically, the application of a specific potential at the bioanode has been shown to stimulate the formation of an electro-active biofilm, but the underlying mechanisms remain poorly understood. In this study, we have investigated the effect of an applied potential on the formation and electroactivity of biofilms established by Shewanella oneidensis bacteria on graphite felt electrodes in single- and double-chamber reactor configurations in oxic conditions. Using amperometry, cyclic voltammetry, and OCP/Power/Polarization curves techniques, we showed that a potential ranging between -0.3V and +0.5V (vs. Ag/AgCl/KCl sat.) and its converse application to a couple of electrodes leads to different electrochemical behaviors, anodic currents and biofilm architectures. For example, when the bacteria were confined in the anodic compartment of a double-chamber cell, a negative applied potential (-0.3V) at the bioanode favors a mediated electron transfer correlated with the progressive formation of a biofilm that fills the felt porosity and bridges the graphite fibers. In contrast, a positive applied potential (+0.3V) at the bioanode stimulates a direct electron transfer resulting in the fast-bacterial colonization of the fibers only. These results provide significant insight for the understanding of the complex bacteria-electrode interactions in microbial fuel cells. Copyright © 2017 Elsevier B.V. All rights reserved.

  3. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

    Lotts, A.L.; Homan, F.J.; Balthesen, E.; Turner, R.F.

    1977-01-01

    Significant advances have occurred in the development of HTGR fuel and fuel cycle. These accomplishments permit a wide choice of fuel designs, reactor concepts, and fuel cycles. Fuels capable of providing helium outlet temperatures of 750 0 C are available, and fuels capable of 1000 0 C outlet temperatures may be expected from extension of present technology. Fuels have been developed for two basic HTGR designs, one using a spherical (pebble bed) element and the other a prismatic element. Within each concept a number of variations of geometry, fuel composition, and structural materials are permitted. Potential fuel cycles include both low-enriched and high-enriched Th- 235 U, recycle Th- 233 U, and Th-Pu or U-Pu cycles. This flexibility offered by the HTGR is of great practical benefit considering the rapidly changing economics of power production. The inflation of ore prices has increased optimum conversion ratios, and increased the necessity of fuel recycle at an early date. Fuel element makeup is very similar for prismatic and spherical designs. Both use spherical fissile and fertile particles coated with combinations of pyrolytic carbon and silicon carbide. Both use carbonaceous binder materials, and graphite as the structural material. Weak-acid resin (WAR) UO 2 -UC 2 fissile fuels and sol-gel-derived ThO 2 fertile fuels have been selected for the Th- 233 U cycle in the prismatic design. Sol-gel-derived UO 2 UC 2 is the reference fissile fuel for the low-enriched pebble bed design. Both the United States and Federal Republic of Germany are developing technology for fuel cycle operations including fabrication, reprocessing, refabrication, and waste handling. Feasibility of basic processes has been established and designs developed for full-scale equipment. Fuel and fuel cycle technology provide the basis for a broad range of applications of the HTGR. Extension of the fuels to higher operating temperatures and development and commercial demonstration of fuel

  4. Electrocatalytic properties of graphite nanofibers-supported platinum catalysts for direct methanol fuel cells.

    Science.gov (United States)

    Park, Soo-Jin; Park, Jeong-Min; Seo, Min-Kang

    2009-09-01

    Graphite nanofibers (GNFs) treated at various temperatures were used as carbon supports to improve the efficiency of PtRu catalysts. The electrochemical properties of the PtRu/GNFs catalysts were then investigated to evaluate their potential for application in DMFCs. The results indicated that the particle size and dispersibility of PtRu in the catalysts were changed by heat treatment, and the electrochemical activity of the catalysts was improved. Consequently, it was found that heat treatments could have an influence on the surface and structural properties of GNFs, resulting in enhancing an electrocatalytic activity of the catalysts for DMFCs.

  5. The importance of the AVR pebble-bed reactor for the future of nuclear power

    International Nuclear Information System (INIS)

    Pohl, P.

    2006-01-01

    The AVR pebble-bed high temperature gas-cooled reactor (HTGR) at Juelich (Germany)) operated from 1967 to 1988 and was certainly the most important HTGR project of the past. The reactor was the mass test bed for all development steps of HTGR pebble fuel. Some early fuel charges failed under high temperature conditions and contaminated the reactor. An accurate pebble measurement (Cs 137) allowed to clean the core from unwanted pebbles after 1981. The coolant activity went down and remained very low for the remaining reactor operation. A melt-wire experiment in 1986 revealed max. coolant temperatures of >1280 deg. C and fuel temperatures of >1350 deg. C, explained by under-estimated bypasses. The fuel still in the core achieved high burn-ups and showed under the extreme temperature conditions excellent fission product retention. Thus, the AVR operation qualified the HTGR fuel, and an average discharge burn-up of 112% fifa revealed an excellent fuel economy of the pebble-bed reactor. Furthermore, the AVR operation offers many meaningful data for code-to-experiment comparisons. (authors)

  6. Preliminary neutronic design of high burnup OTTO cycle pebble bed reactor

    International Nuclear Information System (INIS)

    Setiadipura, T.; Zuhair; Irwanto, D.

    2015-01-01

    The pebble bed type High Temperature Gas-cooled Reactor (HTGR) is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR) which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO) cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM) loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble. (author)

  7. Preliminary Neutronic Design of High Burnup OTTO Cycle Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    T. Setiadipura

    2015-04-01

    Full Text Available The pebble bed type High Temperature Gas-cooled Reactor (HTGR is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble

  8. US/FRG umbrella agreement for cooperation in GCR Development. Fuel, fission products, and graphite subprogram. Quarterly status report, July 1, 1982-September 30, 1982

    International Nuclear Information System (INIS)

    Turner, R.F.

    1982-10-01

    This report describes the status of the cooperative work being performed in the Fuel, Fission Product, and Graphite Subprogram under the HTR-Implementing Agreement of the United States/Federal Republic of Germany Umbrella Agreement for Cooperation in GCR Development. The status is described relative to the commitments in the Subprogram Plan for Fuel, Fission Products, and Graphite, Revision 5, April 1982. The work described was performed during the period July 1, 1982 through September 30, 1982 in the HTGR Base Technology Program at Oak Ridge National Laboratory, the HTGR Fuel and Plant Technology Programs at General Atomic Company (GA), and the Project HTR-Brennstoffkreislauf of the Entwicklungsgemeinschaft HTR at KFA Julich, HRB Mannheim, HOBEG Hanau, and SIGRI Meitingen. The requirement for and format of this quarterly status report are specified in the HTR Implementing Agreement procedures for cooperation. Responsibility for preparation of the quarterly report alternates between GA and KFA

  9. A scaled experimental study of control blade insertion dynamics in Pebble-Bed Fluoride-Salt-Cooled High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Buster, Grant C., E-mail: grant.buster@gmail.com; Laufer, Michael R.; Peterson, Per F.

    2016-07-15

    Highlights: • A granular dynamics scaling methodology is discussed. • Control blade insertion in a representative pebble-bed core is experimentally studied. • Control blade insertion forces and pebble displacements are experimentally measured. • X-ray tomography techniques are used to observe pebble displacement distributions. - Abstract: Direct control element insertion into a pebble-bed reactor core is proposed as a viable control system in molten-salt-cooled pebble-bed reactors. Unlike helium-cooled pebble-bed reactors, this reactor type uses spherical fuel elements with near-neutral buoyancy in the molten-salt coolant, thus reducing contact forces on the fuel elements. This study uses the X-ray Pebble Bed Recirculation Experiment facility to measure the force required to insert a control element directly into a scaled pebble-bed. The required control element insertion force, and therefore the contact force on fuel elements, is measured to be well below recommended limits. Additionally, X-ray tomography is used to observe how the direct insertion of a control element physically displaces spherical fuel elements. The tomography results further support the viability of direct control element insertion into molten-salt-cooled pebble-bed reactor cores.

  10. 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

  11. Advanced High-Temperature Reactor for Production of Electricity and Hydrogen: Molten-Salt-Coolant, Graphite-Coated-Particle-Fuel

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    2002-01-01

    The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-temperature gas-cooled reactors (MHTGRs), such as the General Atomics gas turbine-modular helium reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced reactor concept that operates at essentially ambient pressures and at very high temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large reactor sizes [∼1000 Mw(e)] with passive safety systems to provide the potential for improved economics

  12. Automated Design and Optimization of Pebble-bed Reactor Cores

    International Nuclear Information System (INIS)

    Gougar, Hans D.; Ougouag, Abderrafi M.; Terry, William K.

    2010-01-01

    We present a conceptual design approach for high-temperature gas-cooled reactors using recirculating pebble-bed cores. The design approach employs PEBBED, a reactor physics code specifically designed to solve for and analyze the asymptotic burnup state of pebble-bed reactors, in conjunction with a genetic algorithm to obtain a core that maximizes a fitness value that is a function of user-specified parameters. The uniqueness of the asymptotic core state and the small number of independent parameters that define it suggest that core geometry and fuel cycle can be efficiently optimized toward a specified objective. PEBBED exploits a novel representation of the distribution of pebbles that enables efficient coupling of the burnup and neutron diffusion solvers. With this method, even complex pebble recirculation schemes can be expressed in terms of a few parameters that are amenable to modern optimization techniques. With PEBBED, the user chooses the type and range of core physics parameters that represent the design space. A set of traits, each with acceptable and preferred values expressed by a simple fitness function, is used to evaluate the candidate reactor cores. The stochastic search algorithm automatically drives the generation of core parameters toward the optimal core as defined by the user. The optimized design can then be modeled and analyzed in greater detail using higher resolution and more computationally demanding tools to confirm the desired characteristics. For this study, the design of pebble-bed high temperature reactor concepts subjected to demanding physical constraints demonstrated the efficacy of the PEBBED algorithm.

  13. Biodegradation of phenol in batch and continuous flow microbial fuel cells with rod and granular graphite electrodes.

    Science.gov (United States)

    Moreno, Lyman; Nemati, Mehdi; Predicala, Bernardo

    2018-01-01

    Phenol biodegradation was evaluated in batch and continuous flow microbial fuel cells (MFCs). In batch-operated MFCs, biodegradation of 100-1000 mg L -1 phenol was four to six times faster when graphite granules were used instead of rods (3.5-4.8 mg L -1  h -1 vs 0.5-0.9 mg L -1  h -1 ). Similarly maximum phenol biodegradation rates in continuous MFCs with granular and single-rod electrodes were 11.5 and 0.8 mg L -1  h -1 , respectively. This superior performance was also evident in terms of electrochemical outputs, whereby continuous flow MFCs with granular graphite electrodes achieved maximum current and power densities (3444.4 mA m -3 and 777.8 mW m -3 ) that were markedly higher than those with single-rod electrodes (37.3 mA m -3 and 0.8 mW m -3 ). Addition of neutral red enhanced the electrochemical outputs to 5714.3 mA m -3 and 1428.6 mW m -3 . Using the data generated in the continuous flow MFC, biokinetic parameters including μ m , K S , Y and K e were determined as 0.03 h -1 , 24.2 mg L -1 , 0.25 mg cell (mg phenol) -1 and 3.7 × 10 -4  h -1 , respectively. Access to detailed kinetic information generated in MFC environmental conditions is critical in the design, operation and control of large-scale treatment systems utilizing MFC technology.

  14. 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)

  15. Characteristic behaviour of Pebble Bed High Temperature Gas-cooled Reactors during water ingress events

    International Nuclear Information System (INIS)

    Khoza, Samukelisiwe N.; Serfontein, Dawid E.; Reitsma, Frederik

    2014-01-01

    in order to curb the increase in the multiplication factor, the level of under-moderation in the fuel can be decreased by reducing the heavy metal loading of the fuel spheres and/or the porosity of the graphite reflector block must be increased in order to increase water ingress into the reflectors. For the PBMR-400 the diameter of the central reflector can also be reduced. Results from past studies of water ingress into Pebble Bed Reactors were used to validate and verify the present simulation approach and results. (author)

  16. Optimization of a radially cooled pebble bed reactor - HTR2008-58117

    International Nuclear Information System (INIS)

    Boer, B.; Kloosterman, J. L.; Lathouwers, D.; Van Der Hagen, T. H. J. J.; Van Dam, H.

    2008-01-01

    By altering the coolant flow direction in a pebble bed reactor from axial to radial, the pressure drop can be reduced tremendously. In this case the coolant flows from the outer reflector through the pebble bed and finally to flow paths in the inner reflector. As a consequence, the fuel temperatures are elevated due to the reduced heat transfer of the coolant. However, the power profile and pebble size in a radially cooled pebble bed reactor can be optimized to achieve lower fuel temperatures than current axially cooled designs, while the low pressure drop can be maintained. The radial power profile in the core can be altered by adopting multi-pass fuel management using several radial fuel zones in the core. The optimal power profile yielding a flat temperature profile is derived analytically and is approximated by radial fuel zoning. In this case, the pebbles pass through the outer region of the core first and each consecutive pass is located in a fuel zone closer to the inner reflector. Thereby, the resulting radial distribution of the fissile material in the core is influenced and the temperature profile is close to optimal. The fuel temperature in the pebbles can be further reduced by reducing the standard pebble diameter from 6 cm to a value as low as I cm. An analytical investigation is used to demonstrate the effects on the fuel temperature and pressure drop for both radial and axial cooling. Finally, two-dimensional numerical calculations were performed, using codes for neutronics, thermal-hydraulics and fuel depletion analysis, in order to validate the results for the optimized design that were obtained from the analytical investigations. It was found that for a radially cooled design with an optimized power profile and reduced pebble diameter (below 3.5 cm) both a reduction in the pressure drop (Δp = -2.6 bar), which increases the reactor efficiency with several percent, and a reduction in the maximum fuel temperature (ΔT = -50 deg. C) can be achieved

  17. Direct methanol fuel cell with extended reaction zone anode: PtRu and PtRuMo supported on graphite felt

    Science.gov (United States)

    Bauer, Alex; Gyenge, Előd L.; Oloman, Colin W.

    Pressed graphite felt (thickness ∼350 μm) with electrodeposited PtRu (43 g m -2, 1.4:1 atomic ratio) or PtRuMo (52 g m -2, 1:1:0.3 atomic ratio) nanoparticle catalysts was investigated as an anode for direct methanol fuel cells. At temperatures above 333 K the fuel cell performance of the PtRuMo catalyst was superior compared to PtRu. The power density was 2200 W m -2 with PtRuMo at 5500 A m -2 and 353 K while under the same conditions PtRu yielded 1925 W m -2. However, the degradation rate of the Mo containing catalyst formulation was higher. Compared to conventional gas diffusion electrodes with comparable PtRu catalyst composition and load, the graphite felt anodes gave higher power densities mainly due to the extended reaction zone for methanol oxidation.

  18. Diffusivities of Ag, Cs, Sr, and Kr in TRISO fuel particles and graphite

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Tri-structural isotropic (TRISO) coated particles have been developed and studied since the late 1950s when the concept of coated particles was invented by Roy Huddle of the United Kingdom Atomic Energy Authority. Several decades of work by half a dozen countries on fission product transport in TRISO fuel through numerous irradiation and heating experiments have led to several recommendations of transport data and to the adoption of various sets of diffusion coefficients. In 1997, the International Atomic Energy Agency (IAEA) gathered all these historical results and issued a technical document (TECDOC-978 [IAEA]) that summarizes these sets of recommended diffusion coefficients. Table 1 shows the reference literature articles for the diffusivities that have historically been recommended by the American and German TRISO fuel development programs and that are summarized in the IAEA report (see section 7 for full references of these articles).

  19. A COMPARISON OF PEBBLE MIXING AND DEPLETION ALGORITHMS USED IN PEBBLE-BED REACTOR EQUILIBRIUM CYCLE SIMULATION

    International Nuclear Information System (INIS)

    Gougar, Hans D.; Reitsma, Frederik; Joubert, Wessel

    2009-01-01

    Recirculating pebble-bed reactors are distinguished from all other reactor types by the downward movement through and reinsertion of fuel into the core during operation. Core simulators must account for this movement and mixing in order to capture the physics of the equilibrium cycle core. VSOP and PEBBED are two codes used to perform such simulations, but they do so using different methods. In this study, a simplified pebble-bed core with a specified flux profile and cross sections is used as the model for conducting analyses of two types of burnup schemes. The differences between the codes are described and related to the differences observed in the nuclide densities in pebbles discharged from the core. Differences in the methods for computing fission product buildup and average number densities lead to significant differences in the computed core power and eigenvalue. These test models provide a key component of an overall equilibrium cycle benchmark involving neutron transport, cross section generation, and fuel circulation.

  20. Conceptual design of a passively safe thorium breeder Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Wols, F.J.; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

    2015-01-01

    Highlights: • This work proposes three possible designs for a thorium Pebble Bed Reactor. • A high-conversion PBR (CR > 0.96), passively safe and within practical constraints. • A thorium breeder PBR (220 cm core) in practical regime, but not passively safe. • A passively safe breeder, requiring higher fuel reprocessing and recycling rates. - Abstract: More sustainable nuclear power generation might be achieved by combining the passive safety and high temperature applications of the Pebble Bed Reactor (PBR) design with the resource availability and favourable waste characteristics of the thorium fuel cycle. It has already been known that breeding can be achieved with the thorium fuel cycle inside a Pebble Bed Reactor if reprocessing is performed. This is also demonstrated in this work for a cylindrical core with a central driver zone, with 3 g heavy metal pebbles for enhanced fission, surrounded by a breeder zone containing 30 g thorium pebbles, for enhanced conversion. The main question of the present work is whether it is also possible to combine passive safety and breeding, within a practical operating regime, inside a thorium Pebble Bed Reactor. Therefore, the influence of several fuel design, core design and operational parameters upon the conversion ratio and passive safety is evaluated. A Depressurized Loss of Forced Cooling (DLOFC) is considered the worst safety scenario that can occur within a PBR. So, the response to a DLOFC with and without scram is evaluated for several breeder PBR designs using a coupled DALTON/THERMIX code scheme. With scram it is purely a heat transfer problem (THERMIX) demonstrating the decay heat removal capability of the design. In case control rods cannot be inserted, the temperature feedback of the core should also be able to counterbalance the reactivity insertion by the decaying xenon without fuel temperatures exceeding 1600 °C. Results show that high conversion ratios (CR > 0.96) and passive safety can be combined in

  1. Graphite coated with manganese oxide/multiwall carbon nanotubes composites as anodes in marine benthic microbial fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Fu, Yubin, E-mail: ffyybb@ouc.edu.cn; Yu, Jian; Zhang, Yelong; Meng, Yao

    2014-10-30

    Highlights: • MnO{sub 2}/MWCNTs composites anode exhibits faster reaction kinetics. • The surfaces of MnO{sub 2}/MWCNTs composites anode exhibits better wettability. • A BMFC using the modified anode have excellent power output. - Abstract: Improving anode performance is of great significance to scale up benthic microbial fuel cells (BMFCs) for its marine application to drive oceanography instruments. In this study, manganese oxide (MnO{sub 2})/multiwall carbon nanotubes (MWCNTs) composites are prepared to be as novel anodes in the BMFCs via a direct redox reaction between permanganate ions (MnO{sub 4}{sup −}) and MWCNTs. The results indicate that the MnO{sub 2}/MWCNTs anode has a better wettability, greater kinetic activity and higher power density than that of the plain graphite (PG) anode. It is noted that the MnO{sub 2} (50% weight percent)/MWCNTs anode shows the highest electrochemical performance among them and will be a promising material for improving bioelectricity production of the BMFCs. Finally, a synergistic mechanism of electron transfer shuttle of Mn ions and their redox reactions in the interface between modified anode and bacteria biofilm are proposed to explain its excellent electrochemical performance.

  2. Performances of helium, neon and argon glow discharges for reduction of fuel hydrogen retention in tungsten, stainless steel and graphite

    International Nuclear Information System (INIS)

    Hino, T.; Yamauchi, Y.; Kimura, Y.; Matsumoto, A.; Nishimura, K.; Ueda, Y.

    2012-11-01

    It is quite important to investigate the performance of glow discharge conditionings for controls of in-vessel tritium (T) inventory and hydrogen recycling. For this purpose, first, the deuterium (D) retentions in tungsten (W), graphite (C) and stainless steel (SS) were measured. The retention in W was not small as expected, several times larger than that of SS, although the retention in SS was one order smaller than that of C. Such the large retention in W is owing to the growth of rough surface structure produced by plasma irradiations. For reduction of deuterium retention in W, SS and C, second, inert gas (He, Ne, Ar) glow discharges were conducted under the same condition, and these performances were compared. The removal ratio of deuterium retention was highest in He discharge, and lowest in Ar discharge. These values are well explained by the numerical analyses using SRIM code. The removal ratios for SS and C were significantly large, but quite small for W. This reason is again owing to the rough surface structure in W. For W, thirdly, the hydrogen isotope exchange and the wall baking experiments were conducted. It is found that the wall backing with a temperature higher than 700 K can well reduce the retention, and the hydrogen isotope exchange using deuterium glow discharge is also useful to reduce the tritium retention in the wall. The present results significantly contribute to control the fuel hydrogen retention and to reduce the in-vessel tritium inventory in fusion reactors. (author)

  3. High performance nano-Ni/Graphite electrode for electro-oxidation in direct alkaline ethanol fuel cells

    Science.gov (United States)

    Soliman, Ahmed B.; Abdel-Samad, Hesham S.; Abdel Rehim, Sayed S.; Ahmed, Mohamed A.; Hassan, Hamdy H.

    2016-09-01

    Ni/Graphite electrocatalysts (Ni/G) are successfully prepared through electrodeposition of Ni from acidic (pH = 0.8) and feebly acidic (pH = 5.5) aqueous Ni (II) baths. The efficiencies of such electrodes are investigated as anodes for direct alkaline ethanol fuel cells through their ethanol electrooxidation cyclic voltammetric (CV) response in alkaline medium. A direct proportionality between the amount of the electrodeposited Ni and its CV response is found. The amounts of the deposited Ni from the two baths are recorded using the Electrochemical Quartz Crystal Microbalance (eQCM). The Ni/G electrodes prepared from the feebly acidic bath show a higher electrocatalytic response than those prepared from the acidic bath. Surface morphology of the Ni particles electrodeposited from feebly acidic bath appears in a nano-scale dimension. Various electrochemical experiments are conducted to confirm that the Ni/G ethanol electrooxidation CV response greatly depends on the pH rather than nickel ion concentration of the deposition bath. The eQCM technique is used to detect the crystalline phases of nickel as α-Ni(OH)2/γ-NiOOH and β-Ni(OH)2/β-NiOOH and their in-situ inter-transformations during the potentiodynamic polarization.

  4. Macroporous graphitic carbon foam decorated with polydopamine as a high-performance anode for microbial fuel cell

    Science.gov (United States)

    Jiang, Hongmei; Yang, Lu; Deng, Wenfang; Tan, Yueming; Xie, Qingji

    2017-09-01

    Herein, a macroporous graphitic carbon foam (MGCF) electrode decorated with polydopamine (PDA) is used as a high-performance anode for microbial fuel cell (MFC) applications. The MGCF is facilely prepared by pyrolysis of a powder mixture comprising maltose, nickel nitrate, and ammonia chloride, without using solid porous template. The MGCF is coated with PDA by self-polymerization of dopamine in a basic solution. The MGCF can provide a large surface area for bacterial attachment, and PDA coated on the MGCF electrode can further promote bacterial adhesion resulting from the improved hydrophility, so the MGCF-PDA electrode as an anode in a MFC can show ultrahigh bacterial loading capacity. Moreover, the electrochemical oxidation of flavins at the MGCF-PDA electrode is greatly accelerated, so the extracellular electron transfer mediated by flavins is improved. As a result, the MFC equipped with a MGCF-PDA anode can show a maximum power density of 1735 mW cm-2, which is 6.7 times that of a MFC equipped with a commercial carbon felt anode, indicating a promising anode for MFC applications.

  5. 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.

  6. Special graphites

    International Nuclear Information System (INIS)

    Leveque, P.

    1964-01-01

    A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [fr

  7. Interpretation of measurements made by oscillations of irradiated fuels in natural uranium, graphite-gas piles

    International Nuclear Information System (INIS)

    Laponche, Bernard; Luffin, Jean; Brunet, Max; Guerange, Jacques; Tonolli, Jacky

    1969-06-01

    When considering a pile operation, it is interesting to know the evolution of fuel quality with respect to irradiation, i.e. the variation of its fission rate and of its absorption rate. In order to experimentally obtain these features, a method is to introduce an irradiated cartridge into a critical reactor and to measure the induced effect on its reactivity and on the neutron density at the vicinity of the cartridge. An oscillation method presented in another document and based on a periodic introduction of fuel sample into a critical reactor allows, from the measurement of reactivity variation (global signal), and of the neutron density (local signal), effective macroscopic fission and absorption cross sections of this sample to be obtained. As previous studies revealed that the interpretation of the local signal was notably delicate, this information has been replaced by computed information, the fission rate, which is determined by means of the COREGRAF1 code. Thus, the remaining quantity to be obtained is the fuel absorption rate. The authors report studies performed on several sets of cartridges from different reactors, and with an irradiation range from about 700 to 4000 MWJ/T. In a first part, they describe the characteristics of the studied cartridges, their irradiation and measurement conditions, and the use of the evolution code. In a second part, they try to define the interpretation of oscillation-based measurements by using two methods, a first and fast one which gives an approximation of results, and a more elaborated second one which complies with measurement conditions. The last part presents and discusses the obtained results [fr

  8. CFD study on the supercritical carbon dioxide cooled pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Dali, E-mail: ydlmitd@outlook.com; Peng, Minjun; Wang, Zhongyi

    2015-01-15

    Highlights: • An innovation concept of supercritical carbon dioxide cooled pebble bed reactor is proposed. • Body-centered cuboid (BCCa) arrangement is adopted for the pebbles. • S-CO{sub 2} would be a good candidate coolant for using in pebble bed reactor. - Abstract: The thermal hydraulic study of using supercritical carbon dioxide (S-CO{sub 2}), a superior fluid state brayton cycle medium, in pebble bed type nuclear reactor is assessed through computational fluid dynamics (CFD) methodology. Preliminary concept design of this S-CO{sub 2} cooled pebble bed reactor (PBR) is implemented by the well-known KTA heat transfer correlation and Ergun pressure drop equation. Eddy viscosity transport turbulence model is adopted and verified by KTA calculated results. Distributions of the temperature, velocity, pressure and Nusselt (Nu) number of the coolant near the surface of the middle spherical fuel element are obtained and analyzed. The conclusion of the assessment is that S-CO{sub 2} would be a good candidate coolant for using in pebble bed reactor due primarily to its good heat transfer characteristic and large mass density, which could lead to achieve lower pressure drop and higher power density.

  9. Postirradiation examination of beryllium pebbles

    International Nuclear Information System (INIS)

    Gelles, D.S.

    1998-01-01

    Postirradiation examinations of COBRA-1A beryllium pebbles irradiated in the EBR-II fast reactor at neutron fluences which generated 2700--3700 appm helium have been performed. Measurements included density change, optical microscopy, scanning electron microscopy, and transmission electron microscopy. The major change in microstructure is development of unusually shaped helium bubbles forming as highly non-equiaxed thin platelet-like cavities on the basal plane. Measurement of the swelling due to cavity formation was in good agreement with density change measurements

  10. Recent developments in graphite

    International Nuclear Information System (INIS)

    Cunningham, J.E.

    1983-01-01

    Overall, the HTGR graphite situation is in excellent shape. In both of the critical requirements, fuel blocks and support structures, adequate graphites are at hand and improved grades are sufficiently far along in truncation. In the aerospace field, GraphNOL N3M permits vehicle performance with confidence in trajectories unobtainable with any other existing material. For fusion energy applications, no other graphite can simultaneously withstand both extreme thermal shock and neutron damage. Hence, the material promises to create new markets as well as to offer a better candidate material for existing applications

  11. Contact detection acceleration in pebble flow simulation for pebble bed reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Li, Y.; Ji, W. [Department of Mechanical, Aerospace, and Nuclear Engineering Rensselaer, Polytechnic Institute, 110 8th street, Troy, NY 12180 (United States)

    2013-07-01

    Pebble flow simulation plays an important role in the steady state and transient analysis of thermal-hydraulics and neutronics for Pebble Bed Reactors (PBR). The Discrete Element Method (DEM) and the modified Molecular Dynamics (MD) method are widely used to simulate the pebble motion to obtain the distribution of pebble concentration, velocity, and maximum contact stress. Although DEM and MD present high accuracy in the pebble flow simulation, they are quite computationally expensive due to the large quantity of pebbles to be simulated in a typical PBR and the ubiquitous contacts and collisions between neighboring pebbles that need to be detected frequently in the simulation, which greatly restricted their applicability for large scale PBR designs such as PBMR400. Since the contact detection accounts for more than 60% of the overall CPU time in the pebble flow simulation, the acceleration of the contact detection can greatly enhance the overall efficiency. In the present work, based on the design features of PBRs, two contact detection algorithms, the basic cell search algorithm and the bounding box search algorithm are investigated and applied to pebble contact detection. The influence from the PBR system size, core geometry and the searching cell size on the contact detection efficiency is presented. Our results suggest that for present PBR applications, the bounding box algorithm is less sensitive to the aforementioned effects and has superior performance in pebble contact detection compared with basic cell search algorithm. (authors)

  12. Contact detection acceleration in pebble flow simulation for pebble bed reactor systems

    International Nuclear Information System (INIS)

    Li, Y.; Ji, W.

    2013-01-01

    Pebble flow simulation plays an important role in the steady state and transient analysis of thermal-hydraulics and neutronics for Pebble Bed Reactors (PBR). The Discrete Element Method (DEM) and the modified Molecular Dynamics (MD) method are widely used to simulate the pebble motion to obtain the distribution of pebble concentration, velocity, and maximum contact stress. Although DEM and MD present high accuracy in the pebble flow simulation, they are quite computationally expensive due to the large quantity of pebbles to be simulated in a typical PBR and the ubiquitous contacts and collisions between neighboring pebbles that need to be detected frequently in the simulation, which greatly restricted their applicability for large scale PBR designs such as PBMR400. Since the contact detection accounts for more than 60% of the overall CPU time in the pebble flow simulation, the acceleration of the contact detection can greatly enhance the overall efficiency. In the present work, based on the design features of PBRs, two contact detection algorithms, the basic cell search algorithm and the bounding box search algorithm are investigated and applied to pebble contact detection. The influence from the PBR system size, core geometry and the searching cell size on the contact detection efficiency is presented. Our results suggest that for present PBR applications, the bounding box algorithm is less sensitive to the aforementioned effects and has superior performance in pebble contact detection compared with basic cell search algorithm. (authors)

  13. Dynamics of a small direct cycle pebble bed HTR

    International Nuclear Information System (INIS)

    Verkerk, E.C.; Heek, A.I. van

    2001-01-01

    The Dutch market for combined generation of heat and power identifies a unit size of 40 MW thermal for the conceptual design of a nuclear cogeneration plant. The ACACIA system provides 14 MW(e) electricity combined with 17 t/h of high temperature steam (220 deg. C, 10 bar) with a pebble bed high temperature reactor directly coupled with a helium compressor and a helium turbine. To come to quantitative statements about the ACACIA transient behaviour, a calculational coupling between the high temperature reactor core analysis code package Panthermix (Panther-Thermix/Direkt) and the thermal hydraulic code RELAP5 for the energy conversion system has been made. This paper will present the analysis of safety related transients. The usual incident scenarios Loss of Coolant Incident (LOCI) and Loss of Flow Incident (LOFI) have been analysed. Besides, also a search for the real maximum fuel temperature (inside a fuel pebble anywhere in the core) has been made. It appears that the maximum fuel temperatures are not reached during a LOFI or LOCI with a halted mass flow rate, but for situations with a small mass flow rate, 1-0.5%. As such, a LOFI or LOCI does not represent the worst-case scenario in terms of maximal fuel temperature. (author)

  14. Silver/iron oxide/graphitic carbon composites as bacteriostatic catalysts for enhancing oxygen reduction in microbial fuel cells

    Science.gov (United States)

    Ma, Ming; You, Shijie; Gong, Xiaobo; Dai, Ying; Zou, Jinlong; Fu, Honggang

    2015-06-01

    Biofilms from anode heterotrophic bacteria are inevitably formed over cathodic catalytic sites, limiting the performances of single-chamber microbial fuel cells (MFCs). Graphitic carbon (GC) - based nano silver/iron oxide (AgNPs/Fe3O4/GC) composites are prepared from waste pomelo skin and used as antibacterial oxygen reduction catalysts for MFCs. AgNPs and Fe3O4 are introduced in situ into the composites by one-step carbothermal reduction, enhancing their conductivity and catalytic activity. To investigate the effects of Fe species on the antibacterial and catalytic properties, AgNPs/Fe3O4/GC is washed with sulfuric acid (1 mol L-1) for 0.5 h, 1 h, and 5 h and marked as AgNPs/Fe3O4/GC-x (x = 0.5 h, 1 h and 5 h, respectively). A maximum power density of 1712 ± 35 mW m-2 is obtained by AgNPs/Fe3O4/GC-1 h, which declines by 4.12% after 17 cycles. Under catalysis of all AgNP-containing catalysts, oxygen reduction reaction (ORR) proceeds via the 4e- pathway, and no toxic effects to anode microorganisms result from inhibiting the cathodic biofilm overgrowth. With the exception of AgNPs/Fe3O4/GC-5 h, the AgNPs-containing composites exhibit remarkable power output and coulombic efficiency through lowering proton transfer resistance and air-cathode biofouling. This study provides a perspective for the practical application of MFCs using these efficient antibacterial ORR catalysts.

  15. Behavior of beryllium pebbles under irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Dalle-Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik; Baldwin, D.L.; Gelles, D.S.; Greenwood, L.R.; Kawamura, H.; Oliver, B.M.

    1998-01-01

    Beryllium pebbles are being considered in fusion reactor blanket designs as neutron multiplier. An example is the European `Helium Cooled Pebble Bed Blanket.` Several forms of beryllium pebbles are commercially available but little is known about these forms in response to fast neutron irradiation. Commercially available beryllium pebbles have been irradiated to approximately 1.3 x 10{sup 22} n/cm{sup 2} (E>1 MeV) at 390degC. Pebbles 1-mm in diameter manufactured by Brush Wellman, USA and by Nippon Gaishi Company, Japan, and 3-mm pebbles manufactured by Brush Wellman were included. All were irradiated in the below-core area of the Experimental Breeder Reactor-II in Idaho Falls, USA, in molybdenum alloy capsules containing helium. Post-irradiation results are presented on density change measurements, tritium release by assay, stepped-temperature anneal, and thermal ramp desorption tests, and helium release by assay and stepped-temperature anneal measurements, for Be pebbles from two manufacturing methods, and with two specimen diameters. The experimental results on density change and tritium and helium release are compared with the predictions of the code ANFIBE. (author)

  16. South African safety assessment framework for the pebble bed modular reactor - HTR2008-58192

    International Nuclear Information System (INIS)

    Joubert, J.; Kohtz, N.; Coe, I.

    2008-01-01

    neutronic/thermal hydraulic behaviour of the reactor, of fission product release from the fuel elements, and of activation of fuel matrix and graphite impurities. (authors)

  17. Water-ingress analysis for the 200 MWe pebble-bed modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Zheng Yanhua; Shi Lei; Wang Yan

    2010-01-01

    Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor adopting steam-turbine cycle, which will cause a positive reactivity introduction, as well as the chemical corrosion of graphite fuel elements and reflector structure material. Besides, increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The analysis of such a kind of important and particular accident is significant to verify the inherent safety characteristics of the modular HTR plants. Based on the preliminary design of the 200 MWe high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the design basis accident of a double-ended guillotine break of one heating tube and the beyond design basis accident of a large break of the main steam collection plate have been analyzed by using TINTE code, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature, the primary loop pressure, the graphite corrosion, the water gas releasing amount, as well as the natural convection influence on the condition of failing to close the blower flaps, have been studied in detail. The calculation results indicate that even under some severe hypothetical postulates, the HTR-PM is able to keep the inherent safeties of the modular high temperature gas-cooled reactor and has a relatively good natural plant response, which will not result in environmental radiation hazard.

  18. AGC-3 Graphite Preirradiation Data Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    William Windes; David Swank; David Rohrbaugh; Joseph Lord

    2013-09-01

    This report describes the specimen loading order and documents all pre-irradiation examination material property measurement data for the graphite specimens contained within the third Advanced Graphite Capsule (AGC-3) irradiation capsule. The AGC-3 capsule is third in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens allowing quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades to be generated which will ascertain the in-service behavior of the graphite for pebble bed and prismatic Very High Temperature Reactor (VHTR) designs. The general design of AGC-3 test capsule is similar to the AGC-2 test capsule, material property tests were conducted on graphite specimens prior to loading into the AGC-3 irradiation assembly. However the 6 major nuclear graphite grades in AGC-2 were modified; two previous graphite grades (IG-430 and H-451) were eliminated and one was added (Mersen’s 2114 was added). Specimen testing from three graphite grades (PCEA, 2114, and NBG-17) was conducted at Idaho National Laboratory (INL) and specimen testing for two grades (IG-110 and NBG-18) were conducted at Oak Ridge National Laboratory (ORNL) from May 2011 to July 2013. This report also details the specimen loading methodology for the graphite specimens inside the AGC-3 irradiation capsule. The AGC-3 capsule design requires "matched pair" creep specimens that have similar dose levels above and below the neutron flux profile mid-plane to provide similar specimens with and without an applied load. This document utilized the neutron flux profile calculated for the AGC-3 capsule design, the capsule dimensions, and the size (length) of the selected graphite and silicon carbide samples to create a stacking order that can produce "matched pairs" of graphite samples above and below the AGC-3 capsule elevation mid-point to

  19. Single-phase convection heat transfer characteristics of pebble-bed channels with internal heat generation

    International Nuclear Information System (INIS)

    Meng Xianke; Sun Zhongning; Xu Guangzhan

    2012-01-01

    Graphical abstract: The core of the water-cooled pebble bed reactor is the porous channels which stacked with spherical fuel elements. The gaps between the adjacent fuel elements are complex because they are stochastic and often shift. We adopt electromagnetic induction heating method to overall heat the pebble bed. By comparing and analyzing the experimental data, we get the rule of power distribution and the rule of heat transfer coefficient with particle diameter, heat flux density, inlet temperature and working fluid's Re number. Highlights: ► We adopt electromagnetic induction heating method to overall heat the pebble bed to be the internal heat source. ► The ball diameter is smaller, the effect of the heat transfer is better. ► With Re number increasing, heat transfer coefficient is also increasing and eventually tends to stabilize. ► The changing of heat power makes little effect on the heat transfer coefficient of pebble bed channels. - Abstract: The reactor core of a water-cooled pebble bed reactor includes porous channels that are formed by spherical fuel elements. This structure has notably improved heat transfer. Due to the variability and randomness of the interstices in pebble bed channels, heat transfer is complex, and there are few studies regarding this topic. To study the heat transfer characters of pebble bed channels with internal heat sources, oxidized stainless steel spheres with diameters of 3 and 8 mm and carbon steel spheres with 8 mm diameters are used in a stacked pebble bed. Distilled water is used as a refrigerant for the experiments, and the electromagnetic induction heating method is used to heat the pebble bed. By comparing and analyzing the experimental results, we obtain the governing rules for the power distribution and the heat transfer coefficient with respect to particle diameter, heat flux density, inlet temperature and working fluid Re number. From fitting of the experimental data, we obtain the dimensionless average

  20. Sigma-R: an improved version of the delayed neutron counting device for THTR fuel element verification

    International Nuclear Information System (INIS)

    Galantucci, R.; Haas, R.; Janssens, A.; Agostini, P.; Becker, L.; Bernede, M.; Caldon, L.; Cordani, G.; Guardini, S.

    1989-01-01

    SIGMA is a device designed for the Safeguards verification measurements of the 235 U content of THTR (or AVR) pebble fuel elements; it has been in operation for more than 12 years at the HOBEG fuel fabrication plant in HANAU (FRG). The THTR fuel element is a 60 mm diameter sphere composed of particles of a U-Th mixture in a graphite matrix. The device is schematically composed of an irradiation facility where the pebble is irradiated by 252 Cf and a counting facility where the delayed fission neutrons are monitored. The system has recently been thoroughly updated, the modified instrument being called SIGMA-R. SIGMA-R automatically governs the whole charge-discharge, irradiation and counting sequence, collects the measured counts in the microprocessor memory and performs the relevant calculations, including the statistical evaluation. The instructions are given interactively to the inspector through an IRIS terminal. SIGMA-R provides extensive self diagnosis features. The inspector is also informed explicity of malfunctions during routine operation. The instrument is equipped with a permanent memory which facilitates the inspector's work considerably. SIGMA is complementary to the so-called fuel pebble sampling devices (FPSD) which are designed to take a random sample of 10 pebbles out of 1000 directly from the production line. With SIGMA-R it is now possible to measure the 235 U content of THTR pebbles very accurately. The uncertainty essentially is limited to Poisson statistics and allows us to evaluate the 235 U content in routine operation, with a random error down to 0.15% and a systematic uncertainty estimated at 0.25%, the latter coming from calibration and normalization procedures

  1. Effect of bed configuration on pebble flow uniformity and stagnation in the pebble bed reactor

    International Nuclear Information System (INIS)

    Gui, Nan; Yang, Xingtuan; Tu, Jiyuan; Jiang, Shengyao

    2014-01-01

    Highlights: • Pebble flow uniformity and stagnation characteristics are very important for HTR-PM. • Arc- and brachistochrone-shaped configuration effects are studied by DEM simulation. • Best bed configurations with uniform flow and no stagnated pebbles are suggested. • Detailed quantified characteristics of bed configuration effects are shown for explanation. - Abstract: Pebble flow uniformity and stagnation characteristics are very important for the design of pebble bed high temperature gas-cooled reactor. Pebble flows inside some specifically designed contraction configurations of pebble bed are studied by discrete element method. The results show the characteristics of stagnation rates, recycling rates, radial distribution of pebble velocity and residence time. It is demonstrated clearly that the bed with a brachistochrone-shaped configuration achieves optimum levels of flow uniformity and recycling rate concentration, and almost no pebbles are stagnated in the bed. Moreover, the optimum choice among the arc-shaped bed configurations is demonstrated too. Detailed information shows the quantified characteristics of bed configuration effects on flow uniformity. In addition, a good design of the pebble bed configuration is suggested

  2. Estimation of Graphite Dust Production in ITER TBM

    International Nuclear Information System (INIS)

    Kang, Ji Ho; Kim, Eung Seon

    2013-01-01

    This scheme uses simple equations and the calculation time is much less than others. However, the contact equation requires a specially tuned material properties and instability of system matrix were reported. Second, only a couple of pebbles were modeled using FEM(Finite Element Method) and appropriate boundary and loading conditions are imposed. This scheme gives a detailed information of stress distribution of the pebbles and the stability of calculation is well established. However, the calculation cost is fairly high and only a few pebble can be analyzed in detail at a time with specifically assigned contact conditions. In this study, a prediction model of graphite dust production in ITER(International Thermonuclear Experimental Reactor) TBM(Test Blanket Module) using FEM was introduced and the amount of dust production for an operation cycle was estimated. In this study, graphite dust generation in the reflector zone of ITER TBM was estimated using FE analysis. A unit-cell model was defined to simulate normal contact forces and slip distances on contact points between the center pebble and the surrounding pebbles. The dust production was calculated using Archard equation. The simulation was repeated with different friction coefficient of graphite material to investigate the effect of friction on the dust production. The calculation result showed that the amount of dust production was 2.22∼3.67e-4 g/m 3 which was almost linearly proportional to the friction coefficient of graphite material. The amount of graphite dust production was considered too much small for a dust explosion

  3. Development of a dry-mechanical graphite separation process and elimination of the separated carbon for the reprocessing of spherical HTR fuel elements

    International Nuclear Information System (INIS)

    Kronschnabel, H.

    1982-01-01

    Due to the C-14 distribution the separation of the particle-free outer region of the spherical HTR fuel element with subsequent solidification of the separated carbon makes it possible to reduce by half the remaining C-14 inventory in the inner particle region to be further treated. Separation of the particle-free outer region by a newly developed sphere-peeling milling machine, conditioning the graphite into compacts and in-situ cementation into a salt-mine are the basic elements of this head-end process variation. An annual cavern volume of approx. 2000 m 3 will be needed to ultimately store the graphite of the particle-free outer region, which corresponds to a reprocessing capacity of 50 GWsub(e) installed HTR power. The brush-disintegration of the remaining inner particle region and the resulting peel-brush-preparation are capable of separating 95% of the graphite without any heavy metal losses. With the mentioned reprocessing capacity an annual cavern volume of approx. 16.500 m 3 is required. (orig.) [de

  4. Pebble bed reactor with one-zone core

    International Nuclear Information System (INIS)

    Mueller-Frank, U.; Lohnert, G.

    1977-01-01

    The claim deals with measures to differentiate the flow rate and to remove spherical fuel elements in the core of a pebble bed reactor. Hence the vertical rate of the fuel elements in the border region is for example twice as much as in the centre. A central funnel-shaped outlet on the floor of the core container over which a conical body is placed with its peak pointing upwards, or also the forming of several outlets can be used to adjust to a certain exit rate for the fuel elements. The main target of the invention is a radially extensively constant coolant outlet temperature at the outlet of the core which determines the effectiveness of the connected heat exchanger and thus contributes to economy. (UA) [de

  5. Experimental study on single-phase convection heat transfer characteristics of pebble bed channels with internal heat generation

    International Nuclear Information System (INIS)

    Meng Xianke; Sun Zhongning; Zhou Ping; Xu Guangzhan

    2012-01-01

    The water-cooled pebble bed reactor core is the porous channels stacked with spherical fuel elements, having evident effect on enhancing heat transfer. Owing to the variability and randomness characteristics of it's interstice, pebble bed channels have a very complex heat transfer situation and have little correlative research. In order to research the heat transfer characters of pebble bed channels with internal heat source, electromagnetic induction heating method was adopted for overall heating the pebble bed which was composed of 8 mm diameter steel balls, and the internal heat transfer characteristics were researched. By comparing and analyzing the experimental data, the rule of power distribution and heat transfer coefficient with heat flux density, inlet temperature and working fluid's Re were got. According to the experimental data fitting, the dimensionless average heat transfer coefficient correlation criteria was got. The fitting results are good agreement with the experimental results within 12% difference. (authors)

  6. Comparison of Several Thermal Conductivity Constants for Thermal Hydraulic Calculation of Pebble Bed Reactor

    Science.gov (United States)

    Irwanto, Dwi; Setiadipura, Topan; Pramutadi, Asril

    2017-07-01

    There are two type of High Temperature Gas Reactor (HTGR), prismatic and pebble bed. Pebble Bed type has unique configuration because the fuels are randomly distributed inside the reactor core. In term of safety features, Pebble Bed Reactor (PBR) is one of the most promising reactor type in avoiding severe nuclear accidents. In order to analyze heat transfer and safety of this reactor type, a computer code is now under development. As a first step, calculation method proposed by Stroh [1] is adopted. An approach has been made to treat randomly distributed pebble balls contains fissile material inside the reactor core as a porous medium. Helium gas act as coolant on the reactor system are carrying heat flowing in the area between the pebble balls. Several parameters and constants are taken into account in the new developed code. Progress of the development of the code especially comparison of several thermal conductivity constants for a certain PBR-case are reported in the present study.

  7. Features and validation of discrete element method for simulating pebble flow in reactor core

    International Nuclear Information System (INIS)

    Xu Yong; Li Yanjie

    2005-01-01

    The core of a High-Temperature Gas-cooled Reactor (HTGR) is composed of big number of fuel pebbles, their kinetic behaviors are of great importance in estimating the path and residence time of individual pebble, the evolution of the mixing zone for the assessment of the efficiency of a reactor. Numerical method is highlighted in modern reactor design. In view of granular flow, the Discrete Element Model based on contact mechanics of spheres was briefly described. Two typical examples were presented to show the capability of the DEM method. The former is piling with glass/steel spheres, which provides validated evidences that the simulated angles of repose are in good coincidence with the experimental results. The later is particle discharge in a flat- bottomed silo, which shows the effects of material modulus and demonstrates several features. The two examples show the DEM method enables to predict the behaviors, such as the evolution of pebble profiles, streamlines etc., and provides sufficient information for pebble flow analysis and core design. In order to predict the cyclic pebble flow in a HTGR core precisely and efficiently, both model and code improvement are needed, together with rational specification of physical properties with proper measuring techniques. Strategic and methodological considerations were also discussed. (authors)

  8. Stability and convergence analysis of the quasi-dynamics method for the initial pebble packing

    International Nuclear Information System (INIS)

    Li, Y.; Ji, W.

    2012-01-01

    The simulation for the pebble flow recirculation within Pebble Bed Reactors (PBRs) requires an efficient algorithm to generate an initial overlap-free pebble configuration within the reactor core. In the previous work, a dynamics-based approach, the Quasi-Dynamics Method (QDM), has been proposed to generate densely distributed pebbles in PBRs with cylindrical and annular core geometries. However, the stability and the efficiency of the QDM were not fully addressed. In this work, the algorithm is reformulated with two control parameters and the impact of these parameters on the algorithm performance is investigated. Firstly, the theoretical analysis for a 1-D packing system is conducted and the range of the parameter in which the algorithm is convergent is estimated. Then, this estimation is verified numerically for a 3-D packing system. Finally, the algorithm is applied to modeling the PBR fuel loading configuration and the convergence performance at different packing fractions is presented. Results show that the QDM is efficient in packing pebbles within the realistic range of the packing fraction in PBRs, and it is capable in handling cylindrical geometry with packing fractions up to 63.5%. (authors)

  9. Effect of friction on pebble flow pattern in pebble bed reactor

    International Nuclear Information System (INIS)

    Li, Yu; Gui, Nan; Yang, Xingtuan; Tu, Jiyuan; Jiang, Shengyao

    2016-01-01

    Highlights: • A 3D DEM study on particle–wall/particle friction in pebble bed reactor is carried out. • Characteristic values are defined to evaluate features of pebble flow pattern quantitatively. • Particle–wall friction is dominant to determine flow pattern in a specific pebble bed. • Friction effect of hopper part on flow field is more critical than that of cylinder part. • Three cases of 1:1 full scale practical pebble beds are simulated for demonstration. - Abstract: Friction affects pebble flow pattern in pebble-bed high temperature gas-cooled reactor (HTGR) significantly. Through a series of three dimensional DEM (discrete element method) simulations it is shown that reducing friction can be beneficial and create a uniform and consistent flow field required by nuclear engineering. Particle–wall friction poses a decisive impact on flow pattern, and particle–particle friction usually plays a secondary role; relation between particle–wall friction and flow pattern transition is also concluded. Moreover, new criteria are created to describe flow patterns quantitatively according to crucial issues in HTGR like stagnant zone, radial uniformity and flow sequence. Last but not least, it is proved that friction control of hopper part is more important than that of cylinder part in practical pebble beds, so reducing friction between pebbles and hopper surface is the engineering priority.

  10. Direct methanol fuel cell with extended reaction zone anode: PtRu and PtRuMo supported on graphite felt

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, Alex; Gyenge, Elod L.; Oloman, Colin W. [Department of Chemical and Biological Engineering, The University of British Columbia, 2360 East Mall, Vancouver, BC (Canada)

    2007-05-15

    Pressed graphite felt (thickness {proportional_to}350 {mu}m) with electrodeposited PtRu (43 g m{sup -2}, 1.4:1 atomic ratio) or PtRuMo (52 g m{sup -2}, 1:1:0.3 atomic ratio) nanoparticle catalysts was investigated as an anode for direct methanol fuel cells. At temperatures above 333 K the fuel cell performance of the PtRuMo catalyst was superior compared to PtRu. The power density was 2200 W m{sup -2} with PtRuMo at 5500 A m{sup -2} and 353 K while under the same conditions PtRu yielded 1925 W m{sup -2}. However, the degradation rate of the Mo containing catalyst formulation was higher. Compared to conventional gas diffusion electrodes with comparable PtRu catalyst composition and load, the graphite felt anodes gave higher power densities mainly due to the extended reaction zone for methanol oxidation. (author)

  11. Preparation of anode-electrolyte structures using graphite, sodium bicarbonate or citric acid as pore forming agents for application in solid oxide fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Paz Fiuza, Raigenis da; Silva, Marcos Aurelio da; Guedes, Bruna C.; Pontes, Luiz A.; Boaventura, Jaime Soares [UFBA, Salvador, Bahia (Brazil). Energy and Materials Science Group

    2010-07-01

    Cermets based on Ni supported on YSZ or GDC were prepared for use as anode in direct reform SOFCs. NaHCO3 (Na-Ni-YSZ and Na-Ni-GDC) or citric acid (Ac-Ni-YSZ and Ac-Ni-GDC) were used as pore forming agents (PFAs). The SOFC anode was also prepared using graphite (G-Ni-YSZ and G-Ni-GDC) as PFA for the purposes of comparison. The testing unitary SOFC, planar type, was made by pressing the anode-electrolyte assembly, followed by sintering at 1500 C. After this, LSM (lanthanum and strontium manganite) paint was used for the cathode deposition. The powdered cermets were evaluated in ethanol steam reforming at 650 C. The ethanol conversion was 84% and 32% for cermets Na-Ni-YSZ and G-Ni-YSZ, respectively and the selectivity to H{sub 2} was 32 and 20% for the two cermets, respectively. The Na-Ni-YSZ cermet was ten times more resistant to carbon deposition than the G-Ni-YSZ cermet. SEM micrographs of the anode-electrolyte assembly showed that the use of NaHCO{sub 3} as PFA created a well formed interface between layers with homogeneously distributed pores. In contrast, graphite as PFA formed a loose interface between anode and electrolyte. The performance of the unitary SOFC was evaluated using ethanol, hydrogen or methane as fuel. The cell operated well using any of these fuels; however, they exhibited different electrochemical behavior. (orig.)

  12. Measurement of the enthalpy and specific heat of a Be2C-graphite-UC2 reactor fuel material to 19800K

    International Nuclear Information System (INIS)

    Roth, E.P.

    1980-01-01

    The enthalpy and specific heat of a Be 2 C-graphite-UC 2 composite nuclear fuel material were measured over the temperature range 300 to 1980 0 K using differential scanning calorimetry and liquid argon vaporization calorimetry. The fuel material measured was developed at Sandia National Laboratories for use in pulsed test reactors. The material is a hot-pressed composite consisting of 40 vol % Be 2 C, 49.5 vol % graphite, 3.5 vol % UC 2 and 7.0 vol % void. The specific heat was measured with the differential scanning calorimeter over the temperature range 300 to 950 0 K while the enthalpy was measured over the range 1185 to 1980 0 K with the liquid argon vaporization calorimeter. The normal spectral emittance at a wavelength of 6.5 x 10 -5 cm was measured over the experimental temperature range. The combined experimental enthalpy data were fit using a spline routine and differentiated to give the specific heat. Comparison of the measured specific heat of the composite to the specific heat calculated by summing the contributions of the individual components indicates that the specific heat of the Be 2 C component differs significantly from literature values and is approximately 0.6 cal/g-K (2.5 x 10 3 J/Kg-K) for temperatures above 1000 0 K

  13. Effect of heat source shape on the thermal field in the pebble bed core of High Temperature Gas-cooled Reactor (HTGR)

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Leisheng; Lee, Jaeyoung [Handong Global University, Pohang (Korea, Republic of)

    2015-10-15

    In this study, in order to minimize the error brought by non-uniform heat flux, the spherical heaters are employed as heat source; subsequently, thermal field and heat transfer characteristics of the pebbles are investigated. The thermal field of the pebble surface in PBR is measured with heat source in different shapes. The HTGR design concept exhibits excellent safety features due to the low power density and the large amount of graphite present in the core which gives a large thermal inertia in an accident such as loss of coolant. However, the possible appearance of hot spots in the pebble bed cores of HTGR may affect the integrity of the pebbles, which has drawn the attention of many scientists to investigate the thermal field and to predict the maximum temperature locations in the pebbles using CFD method, Lee et.al has also done some experimental work on measuring the surface temperature of the pebbles as well as visualizing flow patterns of the coolant gas, and it was found that the temperature near the contacting points between pebbles was not higher than the flow stagnation points due to the higher thermal conductivity of the pebble. Certain error of temperature measurement might occur because of not very uniform heat flux in the pebbles since heater in cylindrical shape was utilized as heat source in previous experiment. More uniform heat flux and more complicated thermal profile are found in the result obtained using spherical heaters. The result shows that the temperature in contact point is higher than that in the top point, which is different from the previous results. The complex thermal phenomena observed in the lower-half side-sphere can be explained by the flow pattern near the surface.

  14. Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

    Science.gov (United States)

    Hamzah, Amir; Suwoto; Rohanda, Anis; Adrial, Hery; Bakhri, Syaiful; Sunaryo, Geni Rina

    2018-02-01

    In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

  15. Numerical simulations of helium flow through prismatic fuel elements of very high temperature reactors

    International Nuclear Information System (INIS)

    Ribeiro, Felipe Lopes; Pinto, Joao Pedro C.T.A.

    2013-01-01

    The 4 th generation Very High Temperature Reactor (VHTR) most popular concept uses a graphite-moderated and helium cooled core with an outlet gas temperature of approximately 1000 deg C. The high output temperature allows the use of the process heat and the production of hydrogen through the thermochemical iodine-sulfur process as well as highly efficient electricity generation. There are two concepts of VHTR core: the prismatic block and the pebble bed core. The prismatic block core has two popular concepts for the fuel element: multihole and annular. In the multi-hole fuel element, prismatic graphite blocks contain cylindrical flow channels where the helium coolant flows removing heat from cylindrical fuel rods positioned in the graphite. In the other hand, the annular type fuel element has annular channels around the fuel. This paper shows the numerical evaluations of prismatic multi-hole and annular VHTR fuel elements and does a comparison between the results of these assembly reactors. In this study the analysis were performed using the CFD code ANSYS CFX 14.0. The simulations were made in 1/12 fuel element models. A numerical validation was performed through the energy balance, where the theoretical and the numerical generated heat were compared for each model. (author)

  16. Comparative study of random and uniform models for the distribution of TRISO particles in HTR-10 fuel elements

    International Nuclear Information System (INIS)

    Rosales, J.; Perez, J.; Garcia, C.; Munnoz, A.; Lira, C. A. B. O.

    2015-01-01

    TRISO particles are the specific features of HTR-10 and generally HTGR reactors. Their heterogeneity and random arrangement in graphite matrix of these reactors create a significant modeling challenge. In the simulation of spherical fuel elements using MCNPX are usually created repetitive structures using uniform distribution models. The use of these repetitive structures introduces two major approaches: the non-randomness of the TRISO particles inside the pebbles and the intersection of the pebble surface with the TRISO particles. These approaches could affect significantly the multiplicative properties of the core. In order to study the influence of these approaches in the multiplicative properties was estimated the K inf value in one pebble with white boundary conditions using 4 different configurations regarding the distribution of the TRISO particles inside the pebble: uniform hexagonal model, cubic uniform model, cubic uniform without the effect of cutting and a random distribution model. It was studied the impact these models on core scale solving the problem B1, from the Benchmark Problems presented in a Coordinated Research Program of the IAEA. (Author)

  17. On the Deposition Equilibrium of Carbon Nanotubes or Graphite in the Reforming Processes of Lower Hydrocarbon Fuels

    Directory of Open Access Journals (Sweden)

    Zdzisław Jaworski

    2017-11-01

    Full Text Available The modeling of carbon deposition from C-H-O reformates has usually employed thermodynamic data for graphite, but has rarely employed such data for impure filamentous carbon. Therefore, electrochemical data for the literature on the chemical potential of two types of purified carbon nanotubes (CNTs are included in the study. Parameter values determining the thermodynamic equilibrium of the deposition of either graphite or CNTs are computed for dry and wet reformates from natural gas and liquefied petroleum gas. The calculation results are presented as the atomic oxygen-to-carbon ratio (O/C against temperature (200 to 100 °C for various pressures (1 to 30 bar. Areas of O/C for either carbon deposition or deposition-free are computed, and indicate the critical O/C values below which the deposition can occur. Only three types of deposited carbon were found in the studied equilibrium conditions: Graphite, multi-walled CNTs, and single-walled CNTs in bundles. The temperature regions of the appearance of the thermodynamically stable forms of solid carbon are numerically determined as being independent of pressure and the analyzed reactants. The modeling indicates a significant increase in the critical O/C for the deposition of CNTs against that for graphite. The highest rise in the critical O/C, of up to 290% at 30 bar, was found for the wet reforming process.

  18. Electrochemically assisted organosol method for Pt-Sn nanoparticle synthesis and in situ deposition on graphite felt support: Extended reaction zone anodes for direct ethanol fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Lycke, Derek R.; Gyenge, Elod L. [Department of Chemical and Biological Engineering, The University of British Columbia, 2360 East Mall, Vancouver, BC (Canada)

    2007-03-20

    Two electrochemically assisted variants of the Boenneman organosol method were developed for Pt-Sn nanoparticle synthesis and in situ deposition on graphite felt electrodes (e.g. thickness up to 2 mm). Tetraoctylammonium triethylhydroborate N(C{sub 8}H{sub 17}){sub 4}BH(C{sub 2}H{sub 5}){sub 3} was employed as colloid stabilizer and reductant dissolved in tetrahydrofuran (THF). The role of the electric field at a low deposition current density of 1.25 mA cm{sup -2} was mainly electrophoretic causing the migration and adsorption of N(C{sub 8}H{sub 17}){sub 4}BH(C{sub 2}H{sub 5}){sub 3} on the graphite felt surface where it reduced the PtCl{sub 2}-SnCl{sub 2} mixture. Faradaic electrodeposition was detected mostly for Sn. Typical Pt-Sn loadings were between 0.4 and 0.9 mg cm{sup -2} depending on the type of pre-deposition exposure of the graphite felt: surfactant-adsorption and metal-adsorption variant, respectively. The catalyst surface area and Pt:Sn surface area ratio was determined by anodic striping of an underpotential deposited Cu monolayer. The two deposition variants gave different catalyst surfaces: total area 233 and 76 cm{sup 2} mg{sup -1}, with Pt:Sn surface area ratio of 3.5:1 and 7.7:1 for surfactant and metal adsorption, respectively. Regarding electrocatalysis of ethanol oxidation, voltammetry and chronopotentiometry studies corroborated by direct ethanol fuel cell experiments using 0.5 M H{sub 2}SO{sub 4} as electrolyte, showed that due to a combination of higher catalyst load and Pt:Sn surface ratio, the graphite felt anodes prepared by the metal-adsorption variant gave better performance. The catalyzed graphite felt provided an extended reaction zone for ethanol electrooxidation and it gave higher catalyst mass specific peak power outputs compared to literature data obtained using gas diffusion anodes with carbon black supported Pt-Sn nanoparticles. (author)

  19. Thermo-mechanical screening tests to qualify beryllium pebble beds with non-spherical pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Reimann, Joerg, E-mail: joerg.reimann@partner.kit.edu [IKET, Karlsruhe Institute of Technology, Karlsruhe (Germany); Fretz, Benjamin [KBHF GmbH, Eggenstein-Leopoldshafen (Germany); Pupeschi, Simone [IAM, Karlsruhe Institute of Technology, Karlsruhe (Germany)

    2015-10-15

    Highlights: • In present ceramic breeder blankets, pebble-shaped beryllium is used as a neutron multiplier. • Spherical pebbles are considered as the candidate material, however, non-spherical particles are of economic interest. • Thermo-mechanical pebble bed data do merely exist for non-spherical beryllium grades. • Uniaxial compression tests (UCTs), combined with the Hot Wire Technique (HWT) were used to measure the stress–strain relations and the thermal conductivity. • A small experimental set-up had to be used and a detailed 3D modelling was of prime importance. • Compared to spherical pebble beds, non-spherical pebble beds are generally softer and mainly the thermal conductivity is lower. - Abstract: In present ceramic breeder blankets, pebble-shaped beryllium is used as a neutron multiplier. Fairly spherical pebbles are considered as a candidate material, however, non-spherical particles are of economic interest because production costs are much lower. Yet, thermo-mechanical pebble bed data do merely exist for these beryllium grades, and the blanket relevant potential of these grades cannot be judged. Screening experiments were performed with three different grades of non-spherical beryllium pebbles, produced by different companies, accompanied by experiments with the reference beryllium pebble beds. Uniaxial compression tests (UCTs), combined with the Hot Wire Technique (HWT), were performed to measure both the stress–strain relation and the thermal conductivity, k, at different stress levels. Because of the limited amounts of the non-spherical materials, the experimental set-ups were small and a detailed 3D modelling was of prime importance in order to prove that the used design was appropriate. Compared to the pebble beds consisting of spherical pebbles, non-spherical pebble beds are generally softer (smaller stress for a given strain), and, mainly as a consequence of this, for a given strain value, the thermal conductivity is lower. This

  20. Pebble bed blanket design for deuterium burning tandem mirror reactors

    International Nuclear Information System (INIS)

    Grotz, S.P.; Dhir, V.K.

    1983-01-01

    The UCLA tandem mirror reactor, SATYR, was developed around the capability of tandem mirrors with thermal barriers to burn deuterium at reasonable efficiency levels. The pebble bed concept has been incorporated into our blanket design for the following reasons: 1) Large area-to-volume ratio for purposes of heat removal; 2) Large volume of structure for high thermal capacity thus increasing the safety margin during off-normal incidents; 3) Relatively inexpensive manufacturing costs because of large acceptable tolerances and lack of exotic materials (i.e., lithium). A simplified stress analysis of the blanket module was performed to optimize and simplify the design. The pre-specified stress intensity limitations used were based upon a 30-year predicted lifetime for each module. Along with stress analysis of the vessel a detailed thermal hydraulic analysis of the pebble bed has been completed. Parameters affecting the pebble bed design are fluidization velocity, pressure drop, heat transfer coefficient, thermally induced stress in the spheres and spatial variation of the power density. Although reasonable gross thermal efficiencies of the 2 designs has been achieved (28% for H 2 O and 39% for He) the high net recirculating power fraction for heating and neutral beams results in relatively low net plant efficiencies (21% and 27%). The results show that a blanket can be designed with good thermal efficiency and a relative-ly simple configuration. However, application of this concept to the high Q deuterium-tritium fuel cycle would have difficulties resulting from the need for continuous removal of the tritium. (orig./HP)

  1. 'Once through' cycles in the pebble bed HTR

    International Nuclear Information System (INIS)

    Teuchert, E.

    1977-12-01

    In the pebble bed HTR the 'Once Through' cycles achieve a favorable conservation of uranium resources due to their high burnup and due to the relatively low fissile inventory. A detailed study is given for cycles with highly enriched uranium and thorium, 20% enriched uranium and thorium, and for the low (approximately 8%) enriched cycle. The recommended cycle is based on the known THTR fuel element in the Th/U (93%) cycle. The variant with separate Seed elements and Breed elements presents the best pioneer in view of later recycling and thermal breeding. The minimum proliferation risk is achieved in the Th/U (20%) cycle basing on the fuel element type of the AVR, due to the low amount and high denaturization of the disloaded plutonium. (orig.) [de

  2. AGC-2 Graphite Preirradiation Data Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    William Windes; W. David Swank; David Rohrbaugh; Joseph Lord

    2013-08-01

    This report described the specimen loading order and documents all pre-irradiation examination material property measurement data for the graphite specimens contained within the second Advanced Graphite Capsule (AGC-2) irradiation capsule. The AGC-2 capsule is the second in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens allowing quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades to be generated which will ascertain the in-service behavior of the graphite for pebble bed and prismatic Very High Temperature Reactor (VHTR) designs. Similar to the AGC-1 specimen pre-irradiation examination report, material property tests were conducted on specimens from 18 nuclear graphite types but on an increased number of specimens (512) prior to loading into the AGC-2 irradiation assembly. All AGC-2 specimen testing was conducted at Idaho National Laboratory (INL) from October 2009 to August 2010. This report also details the specimen loading methodology for the graphite specimens inside the AGC-2 irradiation capsule. The AGC-2 capsule design requires “matched pair” creep specimens that have similar dose levels above and below the neutron flux profile mid-plane to provide similar specimens with and without an applied load. This document utilized the neutron flux profile calculated for the AGC-2 capsule design, the capsule dimensions, and the size (length) of the selected graphite and silicon carbide samples to create a stacking order that can produce “matched pairs” of graphite samples above and below the AGC-2 capsule elevation mid-point to provide specimens with similar neutron dose levels.

  3. Criticality calculations of the HTR-10 pebble-bed reactor with SCALE6/CSAS6 and MCNP5

    International Nuclear Information System (INIS)

    Wang, Meng-Jen; Sheu, Rong-Jiun; Peir, Jinn-Jer; Liang, Jenq-Horng

    2014-01-01

    Highlights: • Comparisons of the HTR-10 criticality calculations with SCALE6/CSAS6 and MCNP5 were performed. • The DOUBLEHET unit-cell treatment provides the best k eff estimation among PBR criticality calculations using SCALE6. • The continuous-energy SCALE6 calculations present a non-negligible discrepancy with MCNP5 in three PBR cases. - Abstract: HTR-10 is a 10 MWt prototype pebble-bed reactor (PBR) that presents a doubly heterogeneous geometry for neutronics calculations. An appropriate unit-cell treatment for the associated fuel elements is vital for creating problem-dependent multigroup cross sections. Considering four unit-cell options for resonance self-shielding correction in SCALE6, a series of HTR-10 core models were established using the CSAS6 sequence to systematically investigate how they affected the computational accuracy and efficiency of PBR criticality calculations. Three core configurations, which ranged from simplified infinite lattices to a detailed geometry, were examined. Based on the same ENDF/B-VII.0 cross-section library, multigroup results were evaluated by comparing with continuous-energy SCALE6/CSAS6 and MCNP5 calculations. The comparison indicated that the INFHOMMEDIUM results overestimated the effective multiplication factor (k eff ) by about 2800 pcm, whereas the LATTICECELL and MULTIREGION treatments overestimated k eff values with similar biases at approximately 470–680 pcm. The DOUBLEHET results attained further improvement, reducing the k eff overestimation to approximately 280 pcm. The comparison yielded two unexpected problems from using SCALE6/CSAS6 in HTR-10 criticality calculations. In particular, the continuous-energy CSAS6 calculations in this study present a non-negligible discrepancy with MCNP5, potentially causing a k eff value overestimate of approximately 680 pcm. Notably, using a cell-weighted mixture instead of an explicit model of individual TRISO particles in the pebble fuel zone does not shorten the

  4. Pebble Accretion in Turbulent Protoplanetary Disks

    Science.gov (United States)

    Xu, Ziyan; Bai, Xue-Ning; Murray-Clay, Ruth A.

    2017-09-01

    It has been realized in recent years that the accretion of pebble-sized dust particles onto planetary cores is an important mode of core growth, which enables the formation of giant planets at large distances and assists planet formation in general. The pebble accretion theory is built upon the orbit theory of dust particles in a laminar protoplanetary disk (PPD). For sufficiently large core mass (in the “Hill regime”), essentially all particles of appropriate sizes entering the Hill sphere can be captured. However, the outer regions of PPDs are expected to be weakly turbulent due to the magnetorotational instability (MRI), where turbulent stirring of particle orbits may affect the efficiency of pebble accretion. We conduct shearing-box simulations of pebble accretion with different levels of MRI turbulence (strongly turbulent assuming ideal magnetohydrodynamics, weakly turbulent in the presence of ambipolar diffusion, and laminar) and different core masses to test the efficiency of pebble accretion at a microphysical level. We find that accretion remains efficient for marginally coupled particles (dimensionless stopping time {τ }s˜ 0.1{--}1) even in the presence of strong MRI turbulence. Though more dust particles are brought toward the core by the turbulence, this effect is largely canceled by a reduction in accretion probability. As a result, the overall effect of turbulence on the accretion rate is mainly reflected in the changes in the thickness of the dust layer. On the other hand, we find that the efficiency of pebble accretion for strongly coupled particles (down to {τ }s˜ 0.01) can be modestly reduced by strong turbulence for low-mass cores.

  5. Graphitized nanodiamond supporting PtNi alloy as stable anodic and cathodic electrocatalysts for direct methanol fuel cell

    International Nuclear Information System (INIS)

    Wang, Yongjiao; Zang, Jianbing; Dong, Liang; Pan, Hong; Yuan, Yungang; Wang, Yanhui

    2013-01-01

    Highlights: • The graphitized nanodiamond (GND) showed a higher oxidation-resistance than XC-72. • The PtNi/GND electrocatalytic exhibited greater stability than PtNi/XC-72. • The PtNi/GND had a better catalytic activity for MOR and ORR than Pt/GND. -- Abstract: Surface graphitized nanodiamond (GND) with a diamond core covered by a graphitic carbon shell was prepared by annealing ND at the temperature of 1300 °C in a vacuum of 10 −3 Pa. PtNi electrocatalysts were prepared by a microwave heating polyol method using the prepared GND as a support. The composition and morphology of the PtNi electrocatalysts supported on GND (PtNi/GND) were characterized by X-ray diffraction, transmission electron microscopy and energy dispersion spectra. The results showed that nano-scaled PtNi alloy particles with an atomic ratio of approximately 1:1 were uniformly deposited on the GND through co-reduction process. The electrocatalytic activities of the PtNi/GND electrocatalysts for methanol oxidation reaction (MOR) and oxygen reduction reaction (ORR) were investigated by cyclic voltammetry, chronoamperometry and linear sweep voltammetry. The PtNi/GND exhibited better electrocatalytic activities than the Pt/GND either for MOR and ORR. In comparison with traditional carbon support Vulcan XC-72, GND showed higher oxidation-resistance, and consequently led to greater stability for the PtNi/GND than PtNi/XC-72

  6. Determination of molybdenum, ruthenium, rhodium, and palladium in radioinactive simulated waste of the nuclear fuel cycle by solid sampling graphite furnace atomic absorption spectrometry (GFAAS)

    International Nuclear Information System (INIS)

    Schmiedel, G.; Mainka, E.; Ache, H.J.

    1989-01-01

    In relation with insoluble particles in the nuclear fuel cycle waste, the solid sampling GFAAS was used to determine molybdenum, ruthenium, rhodium, and palladium in such waste. Two methods for the direct determination of these elements are described. The samples must be handled in glove boxes or moreover in hot cells with a robot. The determination of the elements by the cup-in-tube technique needs a very sensitive balance (microbalance) for weighing in μg-range and the handling of this method is not practical in glove boxes and hot cells. An alternative technique of solid sampling GFAAS, which can be used without great problems in glove boxes and hot cells is the slurry technique. In this case two methods have been used. One method uses graphite powder as a diluter, the other is the direct suspension of the sample in a matrix modifier solution. In the case of slurry technique with predilution of the sample with graphite powder, recoveries between 91 and 102% and RSD between 4 and 8% were obtained, whereas in the case of slurry technique with direct suspension of the waste sample recoveries between 91 and 103% and RSD between 14 and 20% for the above mentioned elements were obtained. (orig.)

  7. Uranium-thorium fuel cycle in a very high temperature hybrid system

    International Nuclear Information System (INIS)

    Hernandez, C.R.G.; Oliva, A.M.; Fajardo, L.G.; Garcia, J.A.R.; Curbelo, J.P.; Abadanes, A.

    2011-01-01

    Thorium is a potentially valuable energy source since it is about three to four times as abundant as Uranium. It is also a widely distributed natural resource readily accessible in many countries. Therefore, Thorium fuels can complement Uranium fuels and ensure long term sustainability of nuclear power. The main advantages of the use of a hybrid system formed by a Pebble Bed critical nuclear reactor and two Pebble Bed Accelerator Driven Systems (ADSs) using a Uranium-Thorium (U + Th) fuel cycle are shown in this paper. Once-through and two step U + Th fuel cycle was evaluated. With this goal, a preliminary conceptual design of a hybrid system formed by a Graphite Moderated Gas-Cooled Very High Temperature Reactor and two ADSs is proposed. The main parameters related to the neutronic behavior of the system in a deep burn scheme are optimized. The parameters that describe the nuclear fuel breeding and Minor Actinide stockpile are compared with those of a simple Uranium fuel cycle. (author)

  8. Recompressed exfoliated graphite articles

    Science.gov (United States)

    Zhamu, Aruna; Shi, Jinjun; Guo, Jiusheng; Jang, Bor Z

    2013-08-06

    This invention provides an electrically conductive, less anisotropic, recompressed exfoliated graphite article comprising a mixture of (a) expanded or exfoliated graphite flakes; and (b) particles of non-expandable graphite or carbon, wherein the non-expandable graphite or carbon particles are in the amount of between about 3% and about 70% by weight based on the total weight of the particles and the expanded graphite flakes combined; wherein the mixture is compressed to form the article having an apparent bulk density of from about 0.1 g/cm.sup.3 to about 2.0 g/cm.sup.3. The article exhibits a thickness-direction conductivity typically greater than 50 S/cm, more typically greater than 100 S/cm, and most typically greater than 200 S/cm. The article, when used in a thin foil or sheet form, can be a useful component in a sheet molding compound plate used as a fuel cell separator or flow field plate. The article may also be used as a current collector for a battery, supercapacitor, or any other electrochemical cell.

  9. Fort St. Vrain graphite site mechanical separation concept selection

    International Nuclear Information System (INIS)

    Berry, S.M.

    1993-09-01

    One of the alternatives to the disposal of the Fort St. Vrain (FSV) reactor spent nuclear fuel involves the separation of the fuel rods composed of compacts from the graphite fuel block assembly. After the separation of these two components, the empty graphite fuel blocks would be disposed of as a low level waste (provided the appropriate requirements are met) and the fuel compacts would be treated as high level waste material. This report deals with the mechanical separation aspects concerning physical disassembly of the FSV graphite fuel element into the empty graphite fuel blocks and fuel compacts. This report recommends that a drilling technique is the preferred choice for accessing the, fuel channel holes and that each hole is drilled separately. This report does not cover any techniques or methods to separate the triso fuel particles from the graphite matrix of the fuel compacts

  10. Acceleration of coupled granular flow and fluid flow simulations in pebble bed energy systems

    International Nuclear Information System (INIS)

    Li, Yanheng; Ji, Wei

    2013-01-01

    Highlights: ► Fast simulation of coupled pebble flow and coolant flow in PBR systems is studied. ► Dimension reduction based on axisymmetric geometry shows significant speedup. ► Relaxation of coupling frequency is investigated and an optimal range is determined. ► A total of 80% efficiency increase is achieved by the two fast strategies. ► Fast strategies can be applied to simulating other general fluidized bed systems. -- Abstract: Fast and accurate approaches to simulating the coupled particle flow and fluid flow are of importance to the analysis of large particle-fluid systems. This is especially needed when one tries to simulate pebble flow and coolant flow in Pebble Bed Reactor (PBR) energy systems on a routine basis. As one of the Generation IV designs, the PBR design is a promising nuclear energy system with high fuel performance and inherent safety. A typical PBR core can be modeled as a particle-fluid system with strong interactions among pebbles, coolants and reactor walls. In previous works, the coupled Discrete Element Method (DEM)-Computational Fluid Dynamics (CFD) approach has been investigated and applied to modeling PBR systems. However, the DEM-CFD approach is computationally expensive due to large amounts of pebbles in PBR systems. This greatly restricts the PBR analysis for the real time prediction and inclusion of more physics. In this work, based on the symmetry of the PBR geometry and the slow motion characteristics of the pebble flow, two acceleration strategies are proposed. First, a simplified 3D-DEM/2D-CFD approach is proposed to speed up the DEM-CFD simulation without loss of accuracy. Pebble flow is simulated by a full 3D DEM, while the coolant flow field is calculated with a 2D CFD simulation by averaging variables along the annular direction in the cylindrical and annular geometries. Second, based on the slow motion of pebble flow, the impact of the coupling frequency on the computation accuracy and efficiency is

  11. Acceleration of coupled granular flow and fluid flow simulations in pebble bed energy systems

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yanheng, E-mail: liy19@rpi.edu [Department of Mechanical, Aerospace, and Nuclear Engineering, Rensselaer Polytechnic Institute, 110 8th Street, Troy, NY (United States); Ji, Wei, E-mail: jiw2@rpi.edu [Department of Mechanical, Aerospace, and Nuclear Engineering, Rensselaer Polytechnic Institute, 110 8th Street, Troy, NY (United States)

    2013-05-15

    Highlights: ► Fast simulation of coupled pebble flow and coolant flow in PBR systems is studied. ► Dimension reduction based on axisymmetric geometry shows significant speedup. ► Relaxation of coupling frequency is investigated and an optimal range is determined. ► A total of 80% efficiency increase is achieved by the two fast strategies. ► Fast strategies can be applied to simulating other general fluidized bed systems. -- Abstract: Fast and accurate approaches to simulating the coupled particle flow and fluid flow are of importance to the analysis of large particle-fluid systems. This is especially needed when one tries to simulate pebble flow and coolant flow in Pebble Bed Reactor (PBR) energy systems on a routine basis. As one of the Generation IV designs, the PBR design is a promising nuclear energy system with high fuel performance and inherent safety. A typical PBR core can be modeled as a particle-fluid system with strong interactions among pebbles, coolants and reactor walls. In previous works, the coupled Discrete Element Method (DEM)-Computational Fluid Dynamics (CFD) approach has been investigated and applied to modeling PBR systems. However, the DEM-CFD approach is computationally expensive due to large amounts of pebbles in PBR systems. This greatly restricts the PBR analysis for the real time prediction and inclusion of more physics. In this work, based on the symmetry of the PBR geometry and the slow motion characteristics of the pebble flow, two acceleration strategies are proposed. First, a simplified 3D-DEM/2D-CFD approach is proposed to speed up the DEM-CFD simulation without loss of accuracy. Pebble flow is simulated by a full 3D DEM, while the coolant flow field is calculated with a 2D CFD simulation by averaging variables along the annular direction in the cylindrical and annular geometries. Second, based on the slow motion of pebble flow, the impact of the coupling frequency on the computation accuracy and efficiency is

  12. Artificial graphites

    International Nuclear Information System (INIS)

    Maire, J.

    1984-01-01

    Artificial graphites are obtained by agglomeration of carbon powders with an organic binder, then by carbonisation at 1000 0 C and graphitization at 2800 0 C. After description of the processes and products, we show how the properties of the various materials lead to the various uses. Using graphite enables us to solve some problems, but it is not sufficient to satisfy all the need of the application. New carbonaceous material open application range. Finally, if some products are becoming obsolete, other ones are being developed in new applications [fr

  13. Geochemical fingerprints and pebbles zircon geochronology

    Indian Academy of Sciences (India)

    Home; Journals; Journal of Earth System Science; Volume 125; Issue 7. Geochemical fingerprints and pebbles zircon geochronology: Implications for the provenance and tectonic setting of Lower Cretaceous sediments in the Zhucheng Basin (Jiaodong peninsula, North China). Jin-Long Ni Jun-Lai Liu Xiao-Ling Tang ...

  14. "Smart pebble" designs for sediment transport monitoring

    Science.gov (United States)

    Valyrakis, Manousos; Alexakis, Athanasios; Pavlovskis, Edgars

    2015-04-01

    Sediment transport, due to primarily the action of water, wind and ice, is one of the most significant geomorphic processes responsible for shaping Earth's surface. It involves entrainment of sediment grains in rivers and estuaries due to the violently fluctuating hydrodynamic forces near the bed. Here an instrumented particle, namely a "smart pebble", is developed to investigate the exact flow conditions under which individual grains may be entrained from the surface of a gravel bed. This could lead in developing a better understanding of the processes involved, focusing on the response of the particle during a variety of flow entrainment events. The "smart pebble" is a particle instrumented with MEMS sensors appropriate for capturing the hydrodynamic forces a coarse particle might experience during its entrainment from the river bed. A 3-axial gyroscope and accelerometer registers data to a memory card via a microcontroller, embedded in a 3D-printed waterproof hollow spherical particle. The instrumented board is appropriately fit and centred into the shell of the pebble, so as to achieve a nearly uniform distribution of the mass which could otherwise bias its motion. The "smart pebble" is powered by an independent power to ensure autonomy and sufficiently long periods of operation appropriate for deployment in the field. Post-processing and analysis of the acquired data is currently performed offline, using scientific programming software. The performance of the instrumented particle is validated, conducting a series of calibration experiments under well-controlled laboratory conditions.

  15. Measurements of the purge helium pressure drop across pebble beds packed with lithium orthosilicate and glass pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Abou-Sena, Ali, E-mail: ali.abou-sena@kit.edu; Arbeiter, Frederik; Boccaccini, Lorenzo V.; Schlindwein, Georg

    2014-10-15

    Highlights: • The objective is to measure the purge helium pressure drop across various HCPB-relevant pebble beds packed with lithium orthosilicate and glass pebbles. • The purge helium pressure drop significantly increases with decreasing the pebbles diameter from one run to another. • At the same superficial velocity, the pressure drop is directly proportional to the helium inlet pressure. • The Ergun's equation can successfully model the purge helium pressure drop for the HCPB-relevant pebble beds. • The measured values of the purge helium pressure drop for the lithium orthosilicate pebble bed will support the design of the purge gas system for the HCPB breeder units. - Abstract: The lithium orthosilicate pebble beds of the Helium Cooled Pebble Bed (HCPB) blanket are purged by helium to transport the produced tritium to the tritium extraction system. The pressure drop of the purge helium has a direct impact on the required pumping power and is a limiting factor for the purge mass flow. Therefore, the objective of this study is to measure the helium pressure drop across various HCPB-relevant pebble beds packed with lithium orthosilicate and glass pebbles. The pebble bed was formed by packing the pebbles into a stainless steel cylinder (ID = 30 mm and L = 120 mm); then it was integrated into a gas loop that has four variable-speed side-channel compressors to regulate the helium mass flow. The static pressure was measured at two locations (100 mm apart) along the pebble bed and at inlet and outlet of the pebble bed. The results demonstrated that: (i) the pressure drop significantly increases with decreasing the pebbles diameter, (ii) for the same superficial velocity, the pressure drop is directly proportional to the inlet pressure, and (iii) predictions of Ergun's equation agree well with the experimental results. The measured pressure drop for the lithium orthosilicate pebble bed will support the design of the purge gas system for the HCPB.

  16. A study on Monte Carlo analysis of Pebble-type VHTR core for hydrogen production

    International Nuclear Information System (INIS)

    Kim, Hong Chul

    2005-02-01

    In order to pursue exact the core analysis for VHTR core which will be developed in future, a study on Monte Carol method was carried out. In Korea, pebble and prism type core are under investigation for VHTR core analysis. In this study, pebble-type core was investigated because it was known that it should not only maintain the nuclear fuel integrity but also have the advantage in economical efficiency and safety. The pebble-bed cores of HTR-PROTEUS critical facility in Swiss were selected for the benchmark model. After the detailed MCNP modeling of the whole facility, calculations of nuclear characteristics were performed. The two core configurations, Core 4.3 and Core 5 (reference state no. 3), among the 10 configurations of the HTR-PROTEUS cores were chosen to be analyzed in order to treat different fuel loading pattern and modeled. The former is a random packing core and the latter deterministic packing core. Based on the experimental data and the benchmark result of other research groups for the two different cores, some nuclear characteristics were calculated. Firstly, keff was calculated for these cores. The effect for TRIO homogeneity model was investigated. Control rod and shutdown rod worths also were calculated and the sensitivity analysis on cross-section library and reflector thickness was pursued. Lastly, neutron flux profiles were investigated in reflector regions. It is noted that Monte Carlo analysis of pebble-type VHTR core was firstly carried out in Korea. Also, this study should not only provide the basic data for pebble-type VHTR core analysis for hydrogen production but also be utilized as the verified data to validate a computer code for VHTR core analysis which will be developed in future

  17. "Smart pebble" design for environmental monitoring applications

    Science.gov (United States)

    Valyrakis, Manousos; Pavlovskis, Edgars

    2014-05-01

    Sediment transport, due to primarily the action of water, wind and ice, is one of the most significant geomorphic processes responsible for shaping Earth's surface. It involves entrainment of sediment grains in rivers and estuaries due to the violently fluctuating hydrodynamic forces near the bed. Here an instrumented particle, namely a "smart pebble", is developed to investigate the exact flow conditions under which individual grains may be entrained from the surface of a gravel bed. This could lead in developing a better understanding of the processes involved, while focusing on the response of the particle during a variety of flow entrainment events. The "smart pebble" is a particle instrumented with MEMS sensors appropriate for capturing the hydrodynamic forces a coarse particle might experience during its entrainment from the river bed. A 3-axial gyroscope and accelerometer registers data to a memory card via a microcontroller, embedded in a 3D-printed waterproof hollow spherical particle. The instrumented board is appropriately fit and centred into the shell of the pebble, so as to achieve a nearly uniform distribution of the mass which could otherwise bias its motion. The "smart pebble" is powered by an independent power to ensure autonomy and sufficiently long periods of operation appropriate for deployment in the field. Post-processing and analysis of the acquired data is currently performed offline, using scientific programming software. The performance of the instrumented particle is validated, conducting a series of calibration experiments under well-controlled laboratory conditions. "Smart pebble" allows for a wider range of environmental sensors (e.g. for environmental/pollutant monitoring) to be incorporated so as to extend the range of its application, enabling accurate environmental monitoring which is required to ensure infrastructure resilience and preservation of ecological health.

  18. X-ray tomography investigations on pebble bed structures

    International Nuclear Information System (INIS)

    Reimann, J.; Rolli, R.; Pieritz, R.A.; Ferrero, C.; Di Michiel, M.

    2007-01-01

    Granular materials (pebbles) are used in present ceramic breeder blankets both for the ceramic breeder material and beryllium. The thermal-mechanical behaviour of these pebble beds strongly depends on the arrangement of the pebbles in the bed, their contacts and contact surfaces with other pebbles and with walls. The influence of these quantities is most pronounced for beryllium pebble beds because of the large thermal conductivity ratio of beryllium to helium gas atmosphere. At present, the data base for the pebble bed thermal conductivity (k) and heat transfer coefficient (h) is quite limited for compressed beds and significant discrepancies exist in respect to h. The detailed knowledge of the pebble bed topology is, therefore, essential to better understand the heat transfer mechanisms. In the present work, results from detailed X-ray tomography investigations are reported on pebble topology in i) the pebble bed bulk (which is relevant for k), and ii) the region close to walls with thicknesses of several pebble diameters (relevant for h). At Forschungszentrum Karlsruhe, pebble beds consisting of aluminium spheres with diameters of 2.3 and 5 mm, respectively, (simulating the blanket relevant 1 mm beryllium pebbles), were uniaxially compressed at different pressure levels. High resolution three-dimensional microtomography (MT) experiments were subsequently performed at the European Synchrotron Radiation Facility, Grenoble. Radial and axial void fraction distributions were found to be oscillatory next to the walls and non-oscillatory in the bulk. For non-compressed pebble beds, the bulk void fraction is fairly constant; for compressed beds, a gradient exists along the compression axis. In the bulk, the angular distribution of pebble contacts was found to be fairly constant, indicating that no regular packing structure is induced. In the wall region, the pebble layer touching the wall is composed of zones with hexagonal structures as shown clearly by MT images. This

  19. Graphite Oxidation Simulation in HTR Accident Conditions

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed

    2012-10-19

    Massive air and water ingress, following a pipe break or leak in steam-generator tubes, is a design-basis accident for high-temperature reactors (HTRs). Analysis of these accidents in both prismatic and pebble bed HTRs requires state-of-the-art capability for predictions of: 1) oxidation kinetics, 2) air helium gas mixture stratification and diffusion into the core following the depressurization, 3) transport of multi-species gas mixture, and 4) graphite corrosion. This project will develop a multi-dimensional, comprehensive oxidation kinetics model of graphite in HTRs, with diverse capabilities for handling different flow regimes. The chemical kinetics/multi-species transport model for graphite burning and oxidation will account for temperature-related changes in the properties of graphite, oxidants (O2, H2O, CO), reaction products (CO, CO2, H2, CH4) and other gases in the mixture (He and N2). The model will treat the oxidation and corrosion of graphite in geometries representative of HTR core component at temperatures of 900°C or higher. The developed chemical reaction kinetics model will be user-friendly for coupling to full core analysis codes such as MELCOR and RELAP, as well as computational fluid dynamics (CFD) codes such as CD-adapco. The research team will solve governing equations for the multi-dimensional flow and the chemical reactions and kinetics using Simulink, an extension of the MATLAB solver, and will validate and benchmark the model's predictions using reported experimental data. Researchers will develop an interface to couple the validated model to a commercially available CFD fluid flow and thermal-hydraulic model of the reactor , and will perform a simulation of a pipe break in a prismatic core HTR, with the potential for future application to a pebble-bed type HTR.

  20. Neutron multipilication factors as a function of temperature: a comparison of calculated and measured values for lattices using 233UO2-ThO2 fuel in graphite

    International Nuclear Information System (INIS)

    Newman, D.F.; Gore, B.F.

    1978-01-01

    Neutron multiplication factors calculated as a function of temperature for three graphite-moderated 233 UO 2 -ThO 2 -fueled lattices are correlated with the values measured for these lattices in the high-temperature lattice test reactor (HTLTR). The correlation analysis is accomplished by fitting calculated values of k/sub infinity/(T) to the measured values using two least-squares-fitted correlation coefficients: (a) a normalization factor and (b) a temperature coefficient bias factor. These correlations indicate the existence of a negative (nonconservative) bias in temperature coefficients of reactivity calculated using ENDF/B-IV cross-section data. Use of an alternate cross-section data set for thorium, which has a smaller resonance integral than ENDF/B-IV data, improved the agreement between calculated and measured temperature coefficients of reactivity for the three experimental lattices. The results of the correlations are used to estimate the bias in the temperature coefficient of reactivity calculated for a lattice typical of fresh 233 U recycle fuel for a high-temperature gas-cooled reactor (HTGR). This extrapolation to a lattice having a heavier fissile loading than the experimental lattices is accomplished using a sensitivity analysis of the estimated bias to alternate thorium cross-section data used in calculations of k/sub infinity/(T). The envelope of uncertainty expected to contain the actual values for the temperature coefficient of the reactivity for the 233 U-fueled HTGR lattice studied remains negative at 1600 K (1327 0 C). Although a broader base of experimental data with improved accuracy is always desirable, the existing data base provided by the HTLTR experiments is judged to be adequate for the verification of neutronic calculations for the HTGR containing 233 U fuel at its current state of development

  1. Preliminary Study of 20 MWth Experiment Power Reactor based on Pebble Bed Reactor

    Science.gov (United States)

    Irwanto, Dwi; Permana, Sidik; Pramuditya, Syeilendra

    2017-07-01

    In this study, preliminary design calculations for experimental small power reactor (20 MWt) based on Pebble Bed Reactor (PBR) are performed. PBR technology chosen due to its advantages in neutronic and safety aspects. Several important parameters, such as fissile enrichment, number of fuel passes, burnup and effective multiplication factor are taken into account in the calculation to find neutronic characteristics of the present reactor design.

  2. Pebble bed modular reactor safeguards: developing new approaches and implementing safeguards by design

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Brian David [Los Alamos National Laboratory; Beddingfield, David H [Los Alamos National Laboratory; Durst, Philip [INL; Bean, Robert [INL

    2010-01-01

    The design of the Pebble Bed Modular Reactor (PBMR) does not fit or seem appropriate to the IAEA safeguards approach under the categories of light water reactor (LWR), on-load refueled reactor (OLR, i.e. CANDU), or Other (prismatic HTGR) because the fuel is in a bulk form, rather than discrete items. Because the nuclear fuel is a collection of nuclear material inserted in tennis-ball sized spheres containing structural and moderating material and a PBMR core will contain a bulk load on the order of 500,000 spheres, it could be classified as a 'Bulk-Fuel Reactor.' Hence, the IAEA should develop unique safeguards criteria. In a multi-lab DOE study, it was found that an optimized blend of: (i) developing techniques to verify the plutonium content in spent fuel pebbles, (ii) improving burn-up computer codes for PBMR spent fuel to provide better understanding of the core and spent fuel makeup, and (iii) utilizing bulk verification techniques for PBMR spent fuel storage bins should be combined with the historic IAEA and South African approaches of containment and surveillance to verify and maintain continuity of knowledge of PBMR fuel. For all of these techniques to work the design of the reactor will need to accommodate safeguards and material accountancy measures to a far greater extent than has thus far been the case. The implementation of Safeguards-by-Design as the PBMR design progresses provides an approach to meets these safeguards and accountancy needs.

  3. 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.

  4. Investigation of chemical modifiers for sulfur determination in diesel fuel samples by high-resolution continuum source graphite furnace molecular absorption spectrometry using direct analysis

    Energy Technology Data Exchange (ETDEWEB)

    Huber, Charles S. [Instituto Federal Sul-rio-grandense, Câmpus Pelotas, Pelotas, RS (Brazil); Universidade Federal do Rio Grande do Sul, Instituto de Química, Porto Alegre, RS (Brazil); Vale, Maria Goreti R. [Universidade Federal do Rio Grande do Sul, Instituto de Química, Porto Alegre, RS (Brazil); Instituto Nacional de Ciência e Tecnologia do CNPq, INCT de Energia e Ambiente, Universidade Federal da Bahia, Salvador, BA (Brazil); Welz, Bernhard [Instituto Nacional de Ciência e Tecnologia do CNPq, INCT de Energia e Ambiente, Universidade Federal da Bahia, Salvador, BA (Brazil); Universidade Federal de Santa Catarina, Departamento de Química, Florianópolis, SC (Brazil); Andrade, Jailson B. [Instituto Nacional de Ciência e Tecnologia do CNPq, INCT de Energia e Ambiente, Universidade Federal da Bahia, Salvador, BA (Brazil); Dessuy, Morgana B., E-mail: mbdessuy@ufrgs.br [Universidade Federal do Rio Grande do Sul, Instituto de Química, Porto Alegre, RS (Brazil); Instituto Nacional de Ciência e Tecnologia do CNPq, INCT de Energia e Ambiente, Universidade Federal da Bahia, Salvador, BA (Brazil)

    2015-06-01

    High-resolution continuum source graphite furnace molecular absorption spectrometry has been applied for sulfur determination in diesel fuel. The sharp rotational lines of the carbon monosulfide molecule (formed during the vaporization step) were used to measure the absorbance. The analytical line at 258.056 nm was monitored using the sum of three pixels. Different chemical modifiers were investigated and the mixture of palladium and magnesium was used as chemical modifier in combination with iridium as permanent modifier. L-Cysteine was chosen as sulfur standard and the calibration was done against aqueous standard solutions. The proposed method was applied for the analyses of four diesel samples: two S10 samples and two S500 samples. The trueness of the method was checked with a certified reference material (CRM) of sulfur in diesel fuel (NIST 2724b). Accurate results, for samples and CRM, were achieved after a dilution with propan-1-ol. The following figures of merit were obtained: characteristic mass of 17 ± 3 ng, limit of detection and limit of quantification of 1.4 mg kg{sup −1} and 4.7 mg kg{sup −1}, respectively. - Highlights: • Ir, Ru and Zr were investigated as permanent modifiers. • Ca, Mg, Pd and Pd/Mg were investigated as modifiers in solution. • Indirect determination of sulfur monitoring the molecular absorbance of the CS • Direct analysis of diesel samples using a dilution in propan-1-ol.

  5. Scalable air cathode microbial fuel cells using glass fiber separators, plastic mesh supporters, and graphite fiber brush anodes

    KAUST Repository

    Zhang, Xiaoyuan; Cheng, Shaoan; Liang, Peng; Huang, Xia; Logan, Bruce E.

    2011-01-01

    The combined use of brush anodes and glass fiber (GF1) separators, and plastic mesh supporters were used here for the first time to create a scalable microbial fuel cell architecture. Separators prevented short circuiting of closely

  6. A discrete element method study on the evolution of thermomechanics of a pebble bed experiencing pebble failure

    Energy Technology Data Exchange (ETDEWEB)

    Van Lew, Jon T., E-mail: jtvanlew@fusion.ucla.edu; Ying, Alice; Abdou, Mohamed

    2014-10-15

    The discrete element method (DEM) is used to study the thermal effects of pebble failure in an ensemble of lithium ceramic spheres. Some pebbles crushing in a large system is unavoidable and this study provides correlations between the extent of pebble failure and the reduction in effective thermal conductivity of the bed. In the model, we homogeneously induced failure and applied nuclear heating until dynamic and thermal steady-state. Conduction between pebbles and from pebbles to the boundary is the only mode of heat transfer presently modeled. The effective thermal conductivity was found to decrease rapidly as a function of the percent of failed pebbles in the bed. It was found that the dominant contributor to the reduction was the drop in inter-particle forces as pebbles fail; implying the extent of failure induced may not occur in real pebble beds. The results are meant to assist designers in the fusion energy community who are planning to use packed beds of ceramic pebbles. The evolution away from experimentally measured thermomechanical properties as pebbles fail is necessary for proper operation of fusion reactors.

  7. Methodology of the On-Iine FoIIow Simulation of Pebble-bed High-temperature Reactors

    International Nuclear Information System (INIS)

    Xia Bing; Li Fu; Wei Chunlin; Zheng Yanhua; Chen Fubing; Zhang Jian; Guo Jiong

    2014-01-01

    The on-line fuel management is an essential feature of the pebble-bed high-temperature reactors (PB-HTRs), which is strongly coupled with the normal operation of the reactor. For the purpose of on-line analysis of the continuous shuffling scheme of numerous fuel pebbles, the follow simulation upon the real operation is necessary for the PB-HTRs. In this work, the on-line follow simulation methodology of the PB-HTRs’ operation is described, featured by the parallel treatments of both neutronics analysis and fuel cycling simulation. During the simulation, the operation history of the reactor is divided into a series of burn-up cycles according to the behavior of operation data, in which the steady-state neutron transport equations are solved and the diffusion theory is utilized to determine the physical features of the reactor core. The burn-up equations of heavy metals, fission products and neutron poisons including B-10, decoupled from the pebble flow term, are solved to analyze the burn-up process within a single burn-up cycle. The effect of pebble flow is simulated separately through a discrete fuel shuffling pattern confined by curved pebble flow channels, and the effect of multiple pass of the fuel is represented by logical batches within each spatial region of the core. The on-line thermal-hydraulics feedback is implemented for each bur-up cycle by using the real thermal-hydraulics data of the core operation. The treatment of control rods and absorber balls is carried out by utilizing a coupled neutron transport-diffusion calculation along with discontinuity factors. The physical models mentioned above are established mainly by using a revised version of the V.S.O.P program system. The real operation data of HTR-10 is utilized to verify the methodology presented in this work, which gives good agreement between simulation results and operation data. (author)

  8. High-conversion HTRs and their fuel cycle

    International Nuclear Information System (INIS)

    Gutmann, H.; Hansen, U.; Larsen, H.; Price, M.S.T.

    1976-01-01

    The high-temperature reactors using graphite as structural core material and helium as coolant represent thermal reactor designs with a very high degree of neutron economy which, when using the thorium fuel cycle, offer, at least theoretically, the possibility of thermal breeding. Though this was already known from previous studies, the economic climate at that time was such that the achievement of high conversion ratios conflicted with the incentive for low fuel cycle costs. Consequently, thorium cycle conversion ratios of around 0.6 were found optimum, and the core and fuel element layout followed from the economic ground rules. The recent change in attitude, brought about partly by the slow process of realization of the limits to the earth's accessible high-grade uranium ore resources and more dramatically by the oil crisis, makes it necessary to concentrate attention again on the high conversion fuel cycles. This report discusses the principles of the core design and the fuel cycle layout for High Conversion HTRs (HCHTRs). Though most of the principles apply equally to HTRs of the pebble-bed and the prismatic fuel element design types, the paper concentrates on the latter. Design and fuel cycle strategies for the full utilization of the high conversion potential are compared with others that aim at easier reprocessing and the ''environmental'' fuel cycle. The paper concludes by discussing operating and fuel cycle characteristics and economics of HCHTRs, and how the latter impinge on the allowable price for uranium ore and the available uranium resources. (author)

  9. Innovative approaches to the Management of Irradiated Nuclear Graphite Wastes: Addressing the Challenges through International Collaboration with Project 'GRAPA'

    International Nuclear Information System (INIS)

    Wickham, A.J.; Ojovan, M.; O'Sullivan, P.; )

    2017-01-01

    There exists more than 250.000 tonnes of irradiated (and therefore radioactive) nuclear graphite (i-graphite) in the world, primarily as a result of the development of graphite-moderated power-reactor systems, initially for defence and subsequently for commercial purposes. Only a very small number of such plants have been dismantled and, for most cases, the final destiny of the irradiated graphite remains unresolved. Future high-temperature reactor programmes, such as the Chinese HTR-PM development, will produce more graphite and carbonaceous wastes from both structural components and the fuel pebbles (which are approximately 96% carbonaceous), the latter producing a continuous stream of so-called 'operational waste'. The problem of dismantling irradiated graphite reactor stacks, possibly distorted through neutron damage and in some cases degraded further by radiation-chemical attack by gaseous coolants, and then finding the appropriate treatments and final destiny of the material, has exercised both the European Union and the International Atomic Energy Agency for more than 25 years, seeking to address the different issues and available disposal solutions in different IAEA Member States. An IAEA collaborative research programme on treatment options has recently been completed, and an active group of international specialists in this area has now been established as part of the IAEA International Decommissioning Network under the envelope of Project 'GRAPA' (Irradiated Graphite Processing Approaches), which includes representatives from Belgium, China, France, Germany, India, Italy, Lithuania, Rep. of Korea, Romania, Spain, Switzerland, Ukraine and the Russian Federation with direct responsibilities for various parts of the decommissioning and graphite-disposal process in a variety of reactor designs. Interest has also been expressed by colleagues from Sweden and Japan. Work is in progress on a number of topic areas where weaknesses in the

  10. Study of the Effect of Burnable Poison Particles Applying in a Pebble Bed HTR

    International Nuclear Information System (INIS)

    Wei Chunlin; Zhao Jing; Zhang Jian; Xia Bing

    2014-01-01

    In pebble bed high temperature gas cooled reactors (HTR), spherical fuel elements pass through the core several times to balance the burnup process in the fuel region, resulting in an acceptable shape and peak factor of power density in the simulation analysis. In contrast, when fuel elements pass through the core only once, the peak of power density occurs at the top of the core and its value is too high to be safe. These indicators/parameters can be improved by incorporating burnable poison in the fuel elements under certain conditions. In the current study, burnable poison particles (BPPs) in fuel elements are evaluated. In spite of the strong absorption capability of "1"0B, BPPs can decrease the depletion speed and increase the duration of "1"0B because of the self-shielding effect, resulting in improved shape and peak factor of power distribution. Several BPPs with different radius are discussed in power distribution, following the calculation for a full-scale reactor core with modified VSOP code. According the result, applying BPPs on fuel pebbles is an effective means to improve the distribution of the power density under one-through fuel load in HTR. (author)

  11. Mechanical performance of irradiated beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Dalle-Donne, M.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik

    1998-01-01

    For the Helium Cooled Pebble Bed (HCPB) Blanket, which is one of the two reference concepts studied within the European Fusion Technology Programme, the neutron multiplier consists of a mixed bed of about 2 and 0.1-0.2 mm diameter beryllium pebbles. Beryllium has no structural function in the blanket, however microstructural and mechanical properties are important, as they might influence the material behavior under neutron irradiation. The EXOTIC-7 as well as the `Beryllium` experiments carried out in the HFR reactor in Petten are considered as the most detailed and significant tests for investigating it. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from these irradiation experiments, emphasizing the effects of irradiation of essential material properties and trying to elucidate the processes controlling the property changes. The microstructure, the porosity distribution, the impurity content, the behavior under compression loads and the compatibility of the beryllium pebbles with lithium orthosilicate (Li{sub 4}SiO{sub 4}) during the in-pile irradiation are presented and critically discussed. Qualitative information on ductility and creep obtained by hardness-type measurements are also supplied. (author)

  12. A Preliminary Study of the Effect of Shifts in Packing Fraction on k-effective in Pebble-Bed Reactors

    International Nuclear Information System (INIS)

    Ougouag, Abderrafi Mohammed-El-Ami; Terry, William Knox

    2001-01-01

    A preliminary examination of the effect of pebble packing changes on the reactivity of a pebble-bed reactor (PBR) is performed. As a first step, using the MCNP code, the modeling of a PBR core as a continuous and homogeneous region is compared to the modeling as a collection of discrete pebbles of equal average fuel density. It is shown that the two modeling approaches give the same trends inasmuch as changes in keff are concerned. It is thus shown that for the purpose of identifying trends in keff changes, the use of a homogeneous model is sufficient. A homogeneous model is then used to assess the effect of pebble packing arrangement changes on the reactivity of a PBR core. It is shown that the changes can be large enough to result in prompt criticality. It is shown that for uranium fueled PBRs, thermal feedback could have the potential to offset the increase in activity, whereas for plutonium fueled systems, thermal feedback may not be sufficient for totally offsetting the packing-increase reactivity insertion and could even exacerbate the initial response. It is thus shown that a full study, including reactor kinetics, thermal feedback, and the dynamics of energy deposition and removal is warranted to fully characterize the potential consequences of packing shifts

  13. Installation of a three-dimensional simulation method for core-physical description of pebble bed reactors with multiple recycling process at the example of the AVR

    International Nuclear Information System (INIS)

    Grotkamp, T.

    1984-01-01

    To describe the core-physical behaviour of pebble bed reactors simulation models are used, which reproduce the burn-up/recycling and - resulting - calculate criticality and neutron spectrum as well as neutron flux and temperature distribution. Modelling the AVR-reactor requires a three-dimensional treating for detailed considerations because of the graphite noses extending into the core. Such a system is built up in the present work and compared with the results of the two-dimensional model standardizated from operational side. The agreement is so good that the latter one is sufficient for the calculations accompanying the operation. The comparison with results of measurement is very satisfying in regard to fuel element distribution and temperature coefficient. As in all theoretical investigations there stays a discrepance of a little more than 1 nile against the measurement at the reactivity equivalence of the AVR rod-bank. On the other hand it is possible to reproduce the rod-bank curve resulting of the calibration very exactly with the present model. (orig.) [de

  14. Recent developments concerning French fuel elements used in natural uranium - graphite - CO{sub 2} reactor systems; Developpements recents des elements combustibles francais de la filiere uranium naturel - graphite - CO{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Salesse, M; Stohr, J A; Jeanpierre, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The policy followed in France for the development of fuel elements for reactors belonging to the Electricite de France has been to benefit as much as possible, for each new pile from the most recent technical progress by developing in each case a fuel element allowing the maximum power per channel. The two latest fuel elements thus studied by the French Atomic Energy Commission are of two different types: a tubular uranium element closed at both ends and cooled externally. (This type of element, chosen for the reactors EDF 2, EDF 3 and EDF 4 makes it possible to attain maximum specific powers of the order of 6 MW/metric ton.); an open tubular uranium element cooled both internally and externally, called an annular element which in being studied as a possibility for EDF 5. Such an element makes it possible to attain specific powers of over 12 MW/metric ton. The two types of element have the following common characteristics: - the can, for external cooling, has herning-bone type fins. This type of profile which has been vastly improved recently thereby increasing its thermal efficiency, has the important advantage of avoiding vibration of the element, but has posed problems of resistance to thermal cycling necessitating much research. - the fuel rods are placed inside graphite jackets, this limiting the vertical forces to which they are subjected and protecting them during charging and discharging. On the other hand, these elements present very different problems as for as the following points are concerned: - the characteristics required of the uranium tubes apart of course from a good dimensional stability during irradiation in the two cases are in the case of the closed tubes a very high resistance to external pressure, and in the case of the annular elements a low neutron absorption. Thus for each of these two cases it has been necessary to develop a suitable type of alloy. - a possible loosening of the can during thermal cycling, which is peculiar to the

  15. Ceramic breeder pebble bed packing stability under cyclic loads

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Chunbo, E-mail: chunbozhang@fusion.ucla.edu [Fusion Science and Technology Center, University of California, Los Angeles, CA 90095-1597 (United States); Ying, Alice; Abdou, Mohamed A. [Fusion Science and Technology Center, University of California, Los Angeles, CA 90095-1597 (United States); Park, Yi-Hyun [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    Highlights: • The feasibility of obtaining packing stability for pebble beds is studied. • The responses of pebble bed to cyclic loads have been presented and analyzed in details. • Pebble bed packing saturation and its applications are discussed. • A suggestion is made regarding the improvement of pebbles filling technique. - Abstract: Considering the optimization of blanket performance, it is desired that the bed morphology and packing state during reactor operation are stable and predictable. Both experimental and numerical work are performed to explore the stability of pebble beds, in particular under pulsed loading conditions. Uniaxial compaction tests have been performed for both KIT’s Li{sub 4}SiO{sub 4} and NFRI’s Li{sub 2}TiO{sub 3} pebble beds at elevated temperatures (up to 750 °C) under cyclic loads (up to 6 MPa). The obtained data shows the stress-strain loop initially moves towards the larger strain and nearly saturates after a certain number of cyclic loading cycles. The characterized FEM CAP material models for a Li{sub 4}SiO{sub 4} pebble bed with an edge-on configuration are used to simulate the thermomechanical behavior of pebble bed under ITER pulsed operations. Simulation results have shown the cyclic variation of temperature/stress/strain/gap and also the same saturation trend with experiments under cyclic loads. Therefore, it is feasible for pebble bed to maintain its packing stability during operation when disregarding pebbles’ breakage and irradiation.

  16. Mechanics of a crushable pebble assembly using discrete element method

    International Nuclear Information System (INIS)

    Annabattula, R.K.; Gan, Y.; Zhao, S.; Kamlah, M.

    2012-01-01

    The influence of crushing of individual pebbles on the overall strength of a pebble assembly is investigated using discrete element method. An assembly comprising of 5000 spherical pebbles is assigned with random critical failure energies with a Weibull distribution in accordance with the experimental observation. Then, the pebble assembly is subjected to uni-axial compression (ε 33 =1.5%) with periodic boundary conditions. The crushable pebble assembly shows a significant difference in stress–strain response in comparison to a non-crushable pebble assembly. The analysis shows that a ideal plasticity like behaviour (constant stress with increase in strain) is the characteristic of a crushable pebble assembly with sudden damage. The damage accumulation law plays a critical role in determining the critical stress while the critical number of completely failed pebbles at the onset of critical stress is independent of such a damage law. Furthermore, a loosely packed pebble assembly shows a higher crush resistance while the critical stress is insensitive to the packing factor (η) of the assembly.

  17. A graphite foam reinforced by graphite particles

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, J.J.; Wang, X.Y.; Guo, L.F.; Wang, Y.M.; Wang, Y.P.; Yu, M.F.; Lau, K.T.T. [DongHua University, Shanghai (China). College of Material Science and Engineering

    2007-11-15

    Graphite foam was obtained after carbonization and graphitization of a pitch foam formed by the pyrolysis of coal tar based mesophase pitch mixed with graphite particles in a high pressure and temperature chamber. The graphite foam possessed high mechanical strength and exceptional thermal conductivity after adding the graphite particles. Experimental results showed that the thermal conductivity of modified graphite foam reached 110W/m K, and its compressive strength increased from 3.7 MPa to 12.5 MPa with the addition of 5 wt% graphite particles. Through the microscopic observation, it was also found that fewer micro-cracks were formed in the cell wall of the modified foam as compared with pure graphite foam. The graphitization degree of modified foam reached 84.9% and the ligament of graphite foam exhibited high alignment after carbonization at 1200{sup o}C for 3 h and graphitization at 3000{sup o}C for 10 min.

  18. A Graphite Isotope Ratio Method: A Primer on Estimating Plutonium Production in Graphite Moderated Reactors

    International Nuclear Information System (INIS)

    Gesh, Christopher J.

    2004-01-01

    The Graphite Isotope Ratio Method (GIRM) is a technique used to estimate the total plutonium production in a graphite-moderated reactor. The cumulative plutonium production in that reactor can be accurately determined by measuring neutron irradiation induced isotopic ratio changes in certain impurity elements within the graphite moderator. The method does not require detailed knowledge of a reactor's operating history, although that knowledge can decrease the uncertainty of the production estimate. The basic premise of the Graphite Isotope Ratio Method is that the fluence in non-fuel core components is directly related to the cumulative plutonium production in the nuclear fuel

  19. A simulation of a pebble bed reactor core by the MCNP-4C computer code

    Directory of Open Access Journals (Sweden)

    Bakhshayesh Moshkbar Khalil

    2009-01-01

    Full Text Available Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results, chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.

  20. New sampling method in continuous energy Monte Carlo calculation for pebble bed reactors

    International Nuclear Information System (INIS)

    Murata, Isao; Takahashi, Akito; Mori, Takamasa; Nakagawa, Masayuki.

    1997-01-01

    A pebble bed reactor generally has double heterogeneity consisting of two kinds of spherical fuel element. In the core, there exist many fuel balls piled up randomly in a high packing fraction. And each fuel ball contains a lot of small fuel particles which are also distributed randomly. In this study, to realize precise neutron transport calculation of such reactors with the continuous energy Monte Carlo method, a new sampling method has been developed. The new method has been implemented in the general purpose Monte Carlo code MCNP to develop a modified version MCNP-BALL. This method was validated by calculating inventory of spherical fuel elements arranged successively by sampling during transport calculation and also by performing criticality calculations in ordered packing models. From the results, it was confirmed that the inventory of spherical fuel elements could be reproduced using MCNP-BALL within a sufficient accuracy of 0.2%. And the comparison of criticality calculations in ordered packing models between MCNP-BALL and the reference method shows excellent agreement in neutron spectrum as well as multiplication factor. MCNP-BALL enables us to analyze pebble bed type cores such as PROTEUS precisely with the continuous energy Monte Carlo method. (author)

  1. 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)

  2. Deoxyribonucleic acid directed metallization of platinum nanoparticles on graphite nanofibers as a durable oxygen reduction catalyst for polymer electrolyte fuel cells

    Science.gov (United States)

    Peera, S. Gouse; Sahu, A. K.; Arunchander, A.; Nath, Krishna; Bhat, S. D.

    2015-11-01

    Effective surface functionalization to the hydrophobic graphite nanofibers (GNF) is performed with the biomolecule, namely deoxy-ribo-nucleic-acid (DNA) via π-π interactions. Pt nanoparticles are impregnated on GNF-DNA composite by ethylene glycol reduction method (Pt/GNF-DNA) and its effect on electro catalytic activity for oxygen reduction reaction (ORR) is systemically studied. Excellent dispersion of Pt nanoparticles over GNF-DNA surfaces with no evidence on particle aggregation is a remarkable achievement in this study. This result in higher electro chemical surface area of the catalyst, enhanced ORR behavior with significant enhancement in mass activity. The catalyst is validated in H2-O2 polymer electrolyte fuel cell (PEFC) and a peak power density of 675 mW cm-2 is achieved at a load current density of 1320 mA cm-2 with a minimal catalyst loading of 0.1 mg cm-2 at a cell temperature of 70 °C and 2 bar absolute pressure. Repeated potential cycling up to 10000 cycles in acidic media is also performed for this catalyst and found excellent stability with only 60 mV drop in the ORR half wave potential. The superior behavior of Pt/GNF-DNA catalyst is credited to the robust fibrous structure of GNF and its effective surface functionalization process via π-π interaction.

  3. Development of the loss coefficient correlation for cross flow between graphite fuel blocks in the core of prismatic very high temperature reactor-PMR200

    International Nuclear Information System (INIS)

    Lee, Jeong-Hun; Cho, Hyoung-Kyu; Park, Goon-Cherl

    2016-01-01

    Highlights: • Cross flow experimental data are produced with wedge-shaped and parallel gaps. • The results of a CFD analysis and experimental data are in good agreement. • Pressure loss coefficient for the cross gap between fuel blocks in PMR200 is found. • A new correlation of the cross flow loss coefficient for PMR200 is proposed. - Abstract: The core of the very high temperature reactor (VHTR) PMR200 (a prismatic modular reactor rated at 200 MW of thermal power) consists of hexagonal prismatic fuel blocks and reflector blocks made of graphite. If the core bypass flow ratio increases, the coolant channel flow is decreased and can then lower the heat removal efficiency, resulting in a locally increased fuel block temperature. The coolant channels in the fuel blocks are connected to bypass gaps by the cross gap, complicating flow distribution in the VHTR core. Therefore, reliable estimation of the bypass flow is highly important for the design and safety analysis of the VHTR core. Because of the complexity of the core geometry and gap configuration, it is challenging to predict the flow distribution in the VHTR core. To analyze this flow distribution accurately, it is necessary to determine the cross flow phenomena, and the loss coefficient across the cross gap has to be evaluated to determine the flow distribution in the VHTR core when a lumped parameter code or a flow network analysis code that uses the correlation of the loss coefficient is employed. The purpose of this paper is to develop a loss coefficient correlation applicable to the cross gap in the PMR200 core. The cross flow was evaluated experimentally using the difference between the measured inlet and outlet mass flow rates. Next, the applicability of a commercial computational fluid dynamics (CFD) code, CFX 15, was confirmed by comparing the experimental data and CFD analysis results. To understand the cross flow phenomena, the loss coefficient was evaluated; in the high Reynolds number region

  4. Development of the loss coefficient correlation for cross flow between graphite fuel blocks in the core of prismatic very high temperature reactor-PMR200

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong-Hun, E-mail: huny12@snu.ac.kr; Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr; Park, Goon-Cherl, E-mail: parkgc@snu.ac.kr

    2016-10-15

    Highlights: • Cross flow experimental data are produced with wedge-shaped and parallel gaps. • The results of a CFD analysis and experimental data are in good agreement. • Pressure loss coefficient for the cross gap between fuel blocks in PMR200 is found. • A new correlation of the cross flow loss coefficient for PMR200 is proposed. - Abstract: The core of the very high temperature reactor (VHTR) PMR200 (a prismatic modular reactor rated at 200 MW of thermal power) consists of hexagonal prismatic fuel blocks and reflector blocks made of graphite. If the core bypass flow ratio increases, the coolant channel flow is decreased and can then lower the heat removal efficiency, resulting in a locally increased fuel block temperature. The coolant channels in the fuel blocks are connected to bypass gaps by the cross gap, complicating flow distribution in the VHTR core. Therefore, reliable estimation of the bypass flow is highly important for the design and safety analysis of the VHTR core. Because of the complexity of the core geometry and gap configuration, it is challenging to predict the flow distribution in the VHTR core. To analyze this flow distribution accurately, it is necessary to determine the cross flow phenomena, and the loss coefficient across the cross gap has to be evaluated to determine the flow distribution in the VHTR core when a lumped parameter code or a flow network analysis code that uses the correlation of the loss coefficient is employed. The purpose of this paper is to develop a loss coefficient correlation applicable to the cross gap in the PMR200 core. The cross flow was evaluated experimentally using the difference between the measured inlet and outlet mass flow rates. Next, the applicability of a commercial computational fluid dynamics (CFD) code, CFX 15, was confirmed by comparing the experimental data and CFD analysis results. To understand the cross flow phenomena, the loss coefficient was evaluated; in the high Reynolds number region

  5. Principle design and data of graphite components

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Sumita, Junya; Shibata, Taiju; Iyoku, Tatsuo; Oku, Tatsuo

    2004-01-01

    The High Temperature Engineering Test Reactor (HTTR) constructed by Japan Atomic Energy Research Institute (JAERI) is a graphite-moderated and helium-gas-cooled reactor with prismatic fuel elements of hexagonal blocks. The reactor internal structures of the HTTR are mainly made up of graphite components. As well known, the graphite is a brittle material and there were no available design criteria for brittle materials. Therefore, JAERI had to develop the design criteria taking account of the brittle fracture behavior. In this paper, concept and key specification of the developed graphite design criteria is described, and also an outline of the quality control specified in the design criteria is mentioned

  6. The correction of pebble bed reactor nodal cross sections for the effects of leakage and depletion history

    Science.gov (United States)

    Hudson, Nathanael Harrison

    An accurate and computationally fast method to generate nodal cross sections for the Pebble Bed Reactor (PBR) was presented. In this method, named Spectral History Correction (SHC), a set of fine group microscopic cross section libraries, pre-computed at specified depletion and moderation states, was coupled with the nodal nuclide densities and group bucklings to compute the new fine group spectrum for each node. The relevant fine group cross-section library was then recollapsed to the local broad group cross-section structure with this new fine group spectrum. This library set was tracked in terms of fuel isotopic densities. Fine group modulation factors (to correct the homogeneous flux for heterogeneous effects) and fission spectra were also stored with the cross section library. As the PBR simulation converges to a steady state fuel cycle, the initial nodal cross section library becomes inaccurate due to the burnup of the fuel and the neutron leakage into and out of the node. Because of the recirculation of discharged fuel pebbles with fresh fuel pebbles, a node can consist of a collection of pebbles at various burnup stages. To account for the nodal burnup, the microscopic cross sections were combined with nodal averaged atom densities to approximate the fine group macroscopic cross-sections for that node. These constructed, homogeneous macroscopic cross sections within the node were used to calculate a numerical solution for the fine group spectrum with B1 theory. This new fine spectrum was used to collapse the pre-computed microscopic cross section library to the broad group structure employed by the fuel cycle code. This SHC technique was developed and practically implemented as a subroutine within the PBR fuel cycle code PEBBED. The SHC subroutine was called to recalculate the broad group cross sections during the code convergence. The result was a fast method that compared favorably to the benchmark scheme of cross section calculation with the lattice

  7. Fabrication of modified lithium orthosilicate pebbles by addition of titania

    Energy Technology Data Exchange (ETDEWEB)

    Knitter, R., E-mail: regina.knitter@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-WPT), Karlsruhe, 76021 (Germany); Kolb, M.H.H.; Kaufmann, U. [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-WPT), Karlsruhe, 76021 (Germany); Goraieb, A.A. [Goraieb Versuchstechnik (GVT), Karlsruhe, 76227 (Germany)

    2013-11-15

    Highlights: ► Lithium orthosilicate pebbles with additions of titania were fabricated by a modified melt-based process. ► The fabricated pebbles exhibit a very fine-grained microstructure with lithium metatitanate as a secondary phase. ► Due to the addition of titanate, the crush load of the pebbles was significantly increased. ► The closed porosity was found to be slightly increased with increasing titanate content. -- Abstract: Lithium orthosilicate pebbles are one of the ceramic tritium breeder materials destined for the European solid breeder test blanket modules of ITER, the large-scale scientific experiment intended to prove the viability of fusion as an energy source, presently under construction in Cadarache, France. While the current reference material is fabricated by melt-spraying with 2.5 wt.% excess of silica, resulting in a two-phase material of lithium orthosilicate and metasilicate, a modified melt-based process was used to fabricate breeder pebbles with additions of titania in order to obtain pebbles with lithium metatitanate as a secondary phase. The fabricated two-phase pebbles exhibit a fine-grained microstructure and increased crush loads. The optimum titanate content has yet to be evaluated, nonetheless the pebbles may have the potential to combine the advantages of both lithium orthosilicate and metatitanate breeder ceramics.

  8. Scalable air cathode microbial fuel cells using glass fiber separators, plastic mesh supporters, and graphite fiber brush anodes

    KAUST Repository

    Zhang, Xiaoyuan

    2011-01-01

    The combined use of brush anodes and glass fiber (GF1) separators, and plastic mesh supporters were used here for the first time to create a scalable microbial fuel cell architecture. Separators prevented short circuiting of closely-spaced electrodes, and cathode supporters were used to avoid water gaps between the separator and cathode that can reduce power production. The maximum power density with a separator and supporter and a single cathode was 75±1W/m3. Removing the separator decreased power by 8%. Adding a second cathode increased power to 154±1W/m3. Current was increased by connecting two MFCs connected in parallel. These results show that brush anodes, combined with a glass fiber separator and a plastic mesh supporter, produce a useful MFC architecture that is inherently scalable due to good insulation between the electrodes and a compact architecture. © 2010 Elsevier Ltd.

  9. Thermally exfoliated graphite oxide

    Science.gov (United States)

    Prud'Homme, Robert K. (Inventor); Aksay, Ilhan A. (Inventor); Abdala, Ahmed (Inventor)

    2011-01-01

    A modified graphite oxide material contains a thermally exfoliated graphite oxide with a surface area of from about 300 sq m/g to 2600 sq m/g, wherein the thermally exfoliated graphite oxide displays no signature of the original graphite and/or graphite oxide, as determined by X-ray diffraction.

  10. Progress in radioactive graphite waste management

    International Nuclear Information System (INIS)

    2010-07-01

    Radioactive graphite constitutes a major waste stream which arises during the decommissioning of certain types of nuclear installations. Worldwide, a total of around 250 000 tonnes of radioactive graphite, comprising graphite moderators and reflectors, will require management solutions in the coming years. 14 C is the radionuclide of greatest concern in nuclear graphite; it arises principally through the interaction of reactor neutrons with nitrogen, which is present in graphite as an impurity or in the reactor coolant or cover gas. 3 H is created by the reactions of neutrons with 6 Li impurities in graphite as well as in fission of the fuel. 36 Cl is generated in the neutron activation of chlorine impurities in graphite. Problems in the radioactive waste management of graphite arise mainly because of the large volumes requiring disposal, the long half-lives of the main radionuclides involved and the specific properties of graphite - such as stored Wigner energy, graphite dust explosibility and the potential for radioactive gases to be released. Various options for the management of radioactive graphite have been studied but a generally accepted approach for its conditioning and disposal does not yet exist. Different solutions may be appropriate in different cases. In most of the countries with radioactive graphite to manage, little progress has been made to date in respect of the disposal of this material. Only in France has there been specific thinking about a dedicated graphite waste-disposal facility (within ANDRA): other major producers of graphite waste (UK and the countries of the former Soviet Union) are either thinking in terms of repository disposal or have no developed plans. A conference entitled 'Solutions for Graphite Waste: a Contribution to the Accelerated Decommissioning of Graphite Moderated Nuclear Reactors' was held at the University of Manchester 21-23 March 2007 in order to stimulate progress in radioactive graphite waste management

  11. How cores grow by pebble accretion. I. Direct core growth

    Science.gov (United States)

    Brouwers, M. G.; Vazan, A.; Ormel, C. W.

    2018-03-01

    Context. Planet formation by pebble accretion is an alternative to planetesimal-driven core accretion. In this scenario, planets grow by the accretion of cm- to m-sized pebbles instead of km-sized planetesimals. One of the main differences with planetesimal-driven core accretion is the increased thermal ablation experienced by pebbles. This can provide early enrichment to the planet's envelope, which influences its subsequent evolution and changes the process of core growth. Aims: We aim to predict core masses and envelope compositions of planets that form by pebble accretion and compare mass deposition of pebbles to planetesimals. Specifically, we calculate the core mass where pebbles completely evaporate and are absorbed before reaching the core, which signifies the end of direct core growth. Methods: We model the early growth of a protoplanet by calculating the structure of its envelope, taking into account the fate of impacting pebbles or planetesimals. The region where high-Z material can exist in vapor form is determined by the temperature-dependent vapor pressure. We include enrichment effects by locally modifying the mean molecular weight of the envelope. Results: In the pebble case, three phases of core growth can be identified. In the first phase (Mcore mixes outwards, slowing core growth. In the third phase (Mcore > 0.5M⊕), the high-Z inner region expands outwards, absorbing an increasing fraction of the ablated material as vapor. Rainout ends before the core mass reaches 0.6 M⊕, terminating direct core growth. In the case of icy H2O pebbles, this happens before 0.1 M⊕. Conclusions: Our results indicate that pebble accretion can directly form rocky cores up to only 0.6 M⊕, and is unable to form similarly sized icy cores. Subsequent core growth can proceed indirectly when the planet cools, provided it is able to retain its high-Z material.

  12. A Sub-channel Analysis of a VHTR Fuel Block with Tin Gap-Filler

    International Nuclear Information System (INIS)

    Cho, Chung Ho; Kim, Yong Hee; Yi, Yong Sun; Kim, Hong Pyo

    2005-01-01

    In the Nuclear Hydrogen Development and Demonstration (NHDD) project, two types of VHTRs (Very High Temperature Reactors), prismatic or pebble bed, are under investigation as the nuclear heat source for hydrogen production. In general, the targeted coolant outlet temperature of VHTR is 950∼1000 .deg. C and the maximum allowable fuel temperature is 1250 .deg. C during the normal operation. In the case of the prismatic reactor (PMR), conventional fuel designs result in a small margin in the maximum fuel temperature. This is one of the biggest demerits of the prismatic type In this paper, a technique of lowering the maximum fuel temperature is suggested. The PMR fuel assembly is comprised of many coolant holes and fuel channels. Cylindrical fuel compacts are stacked inside the fuel channel. Consequently, there should be a physical gap between the fuel compact and graphite block, which is filled with the He gas in the conventional design. The heat transfer coefficient of the He gap is very poor, and this increases the fuel temperature substantially. In the proposed design measure, the gap is filled with a liquid metal, tin (Sn) that has a very high thermal conductivity. The effects of tin in the gap with gap distance variation in the viewpoint of thermal hydraulics are quantitatively discussed. Also, the effects of the variations of the axial power distribution are discussed

  13. HTGR Technology Family Assessment for a Range of Fuel Cycle Missions

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Samuel E. Bays; Nick Soelberg

    2010-08-01

    This report examines how the HTGR technology family can provide options for the once through, modified open cycle (MOC), or full recycle fuel cycle strategies. The HTGR can serve all the fuel cycle missions that an LWR can; both are thermal reactors. Additional analyses are warranted to determine if HTGR “full recycle” service could provide improved consumption of transuranic (TRU) material than LWRs (as expected), to analyze the unique proliferation resistance issues associated with the “pebble bed” approach, and to further test and analyze methods to separate TRISO-coated fuel particles from graphite and/or to separate used HTGR fuel meat from its TRISO coating. The feasibility of these two separation issues is not in doubt, but further R&D could clarify and reduce the cost and enable options not adequately explored at present. The analyses here and the now-demonstrated higher fuel burnup tests (after the illustrative designs studied here) should enable future MOC and full recycle HTGR concepts to more rapidly consume TRU, thereby offering waste management advantages. Interest in “limited separation” or “minimum fuel treatment” separation approaches motivates study of impurity-tolerant fuel fabrication. Several issues are outside the scope of this report, including the following: thorium fuel cycles, gas-cooled fast reactors, the reliability of TRISO-coated particles (billions in a reactor), and how soon any new reactor or fuel type could be licensed and then deployed and therefore impact fuel cycle performance measures.

  14. HTGR Technology Family Assessment for a Range of Fuel Cycle Missions

    International Nuclear Information System (INIS)

    Piet, Steven J.; Bays, Samuel E.; Soelberg, Nick

    2010-01-01

    This report examines how the HTGR technology family can provide options for the once through, modified open cycle (MOC), or full recycle fuel cycle strategies. The HTGR can serve all the fuel cycle missions that an LWR can; both are thermal reactors. Additional analyses are warranted to determine if HTGR 'full recycle' service could provide improved consumption of transuranic (TRU) material than LWRs (as expected), to analyze the unique proliferation resistance issues associated with the 'pebble bed' approach, and to further test and analyze methods to separate TRISO-coated fuel particles from graphite and/or to separate used HTGR fuel meat from its TRISO coating. The feasibility of these two separation issues is not in doubt, but further R and D could clarify and reduce the cost and enable options not adequately explored at present. The analyses here and the now-demonstrated higher fuel burnup tests (after the illustrative designs studied here) should enable future MOC and full recycle HTGR concepts to more rapidly consume TRU, thereby offering waste management advantages. Interest in 'limited separation' or 'minimum fuel treatment' separation approaches motivates study of impurity-tolerant fuel fabrication. Several issues are outside the scope of this report, including the following: thorium fuel cycles, gas-cooled fast reactors, the reliability of TRISO-coated particles (billions in a reactor), and how soon any new reactor or fuel type could be licensed and then deployed and therefore impact fuel cycle performance measures.

  15. Prediction of the thermal behavior of a particle spherical fuel element using GITT

    International Nuclear Information System (INIS)

    Pessoa, C.V.; Oliveira, Claudio L. de; Jian, Su

    2008-01-01

    In this work, the transient and steady state heat conduction in a spherical fuel element of a pebble-bed high temperature were studied. This pebble element is composed by a particulate region with spherical inclusions, the fuel UO 2 particles, dispersed in a graphite matrix. A convective heat transfer by helium occurs on the outer surface of the fuel element. The two-energy equation model for the case of pure conduction was applied to this particulate spherical element, generating two macroscopic temperatures, respectively, of the inclusions and of the matrix. The transient analysis was carried out by using the Generalized Integral Transform Technique (GITT) that requires low computational efforts and allows a fast evaluation of the two macroscopic transient temperatures of the particulate region. The solution by GITT leads to a system of ordinary differential equations with the unknown transformed potentials. The mechanical properties (thermal conductivity and specific heat) of the materials were supposed not to depend on the temperature and to be uniform in each region. (author)

  16. The Pebble Bed Modular Reactor: An obituary

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Steve, E-mail: stephen.thomas@gre.ac.u [Public Services International Research Unit (PSIRU), Business School, University of Greenwich, 30 Park Row, London SE10 9LS (United Kingdom)

    2011-05-15

    The High Temperature Gas-cooled Reactor (HTGR) has exerted a peculiar attraction over nuclear engineers. Despite many unsuccessful attempts over half a century to develop it as a commercial power reactor, there is still a strong belief amongst many nuclear advocates that a highly successful HTGR technology will emerge. The most recent attempt to commercialize an HTGR design, the Pebble Bed Modular Reactor (PBMR), was abandoned in 2010 after 12 years of effort and the expenditure of a large amount of South African public money. This article reviews this latest attempt to commercialize an HTGR design and attempts to identify which issues have led to its failure and what lessons can be learnt from this experience. It concludes that any further attempts to develop HTGRs using Pebble Bed technology should only be undertaken if there is a clear understanding of why earlier attempts have failed and a high level of confidence that earlier problems have been overcome. It argues that the PBMR project has exposed serious weaknesses in accountability mechanisms for the expenditure of South African public money. - Research highlights: {yields} In this study we examine the reasons behind the failure of the South African PBMR programme. {yields} The study reviews the technical issues that have arisen and lessons for future reactor developments. {yields} The study also identifies weaknesses in the accountability mechanisms for public spending.

  17. Comparison of stochastic models in Monte Carlo simulation of coated particle fuels

    International Nuclear Information System (INIS)

    Yu Hui; Nam Zin Cho

    2013-01-01

    There is growing interest worldwide in very high temperature gas cooled reactors as candidates for next generation reactor systems. For design and analysis of such reactors with double heterogeneity introduced by the coated particle fuels that are randomly distributed in graphite pebbles, stochastic transport models are becoming essential. Several models were reported in the literature, such as coarse lattice models, fine lattice stochastic (FLS) models, random sequential addition (RSA) models, metropolis models. The principles and performance of these stochastic models are described and compared in this paper. Compared with the usual fixed lattice methods, sub-FLS modeling allows more realistic stochastic distribution of fuel particles and thus results in more accurate criticality calculation. Compared with the basic RSA method, sub-FLS modeling requires simpler and more efficient overlapping checking procedure. (authors)

  18. Treat upgrade fuel fabrication

    International Nuclear Information System (INIS)

    Davidson, K.V.; Schell, D.H.

    1979-01-01

    An extrusion and thermal treatment process was developed to produce graphite fuel rods containing a dispersion of enriched UO 2 . These rods will be used in an upgraded version of the Transient Reactor Test Facility (TREAT). The improved fuel provides a higher graphite matrix density, better fuel dispersion and higher thermal capabilities than the existing fuel

  19. Study of the strength of the internal can for internally and externally cooled fuel elements intended for gas graphite reactors; Etude de la tenue de la gaine interne pour-element combustible a refroidissement interne et externe d'un reacteur graphite-gaz

    Energy Technology Data Exchange (ETDEWEB)

    Boudouresque, B; Courcon, P; Lestiboubois, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    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{sup 2} gas pressure, should remain in contact with the fuel. (authors) [French] La cartouche d'un element combustible annulaire, a refroidissement interne et externe pour reacteur graphite-gaz, est composee d'un tube combustible en uranium, d'une gaine externe et d'une gaine interne en alliage de magnesium. Pour que l'echange thermique entre la gaine interne et le combustible soit bon, il faut que la gaine reste appliquee sur l'uranium quel que soit le regime de temperature. Cette note a pour but de montrer comment, d'apres une etude theorique, le jeu combustible-gaine interne varie au cours des operations de gainage, de chargement dans le reacteur, et des cyclages thermiques. Les parametres suivants sont etudies: diametres de tube, pression du gaz caloporteur, temperature d'entree du gaz, plasticite de l'alliage de gaine. Il est montre que, quel que soit le regime de fonctionnement, la gaine interne d'un element 77 x 95, en projet pour un reacteur graphite-gaz sous pression de 40 kg/cm{sup 2}, doit rester appliquee sur le combustible. (auteurs)

  20. Bridged graphite oxide materials

    Science.gov (United States)

    Herrera-Alonso, Margarita (Inventor); McAllister, Michael J. (Inventor); Aksay, Ilhan A. (Inventor); Prud'homme, Robert K. (Inventor)

    2010-01-01

    Bridged graphite oxide material comprising graphite sheets bridged by at least one diamine bridging group. The bridged graphite oxide material may be incorporated in polymer composites or used in adsorption media.

  1. Direct deterministic method for neutronics analysis and computation of asymptotic burnup distribution in a recirculating pebble-bed reactor

    International Nuclear Information System (INIS)

    Terry, W.K.; Gougar, H.D.; Ougouag, A.M.

    2002-01-01

    A new deterministic method has been developed for the neutronics analysis of a pebble-bed reactor (PBR). The method accounts for the flow of pebbles explicitly and couples the flow to the neutronics. The method allows modeling of once-through cycles as well as cycles in which pebbles are recirculated through the core an arbitrary number of times. This new work is distinguished from older methods by the systematically semi-analytical approach it takes. In particular, whereas older methods use the finite-difference approach (or an equivalent one) for the discretization and the solution of the burnup equation, the present work integrates the relevant differential equation analytically in discrete and complementary sub-domains of the reactor. Like some of the finite-difference codes, the new method obtains the asymptotic fuel-loading pattern directly, without modeling any intermediate loading pattern. This is a significant advantage for the design and optimization of the asymptotic fuel-loading pattern. The new method is capable of modeling directly both the once-through-then-out fuel cycle and the pebble recirculating fuel cycle. Although it currently includes a finite-difference neutronics solver, the new method has been implemented into a modular code that incorporates the framework for the future coupling to an efficient solver such as a nodal method and to modern cross section preparation capabilities. In its current state, the deterministic method presented here is capable of quick and efficient design and optimization calculations for the in-core PBR fuel cycle. The method can also be used as a practical 'scoping' tool. It could, for example, be applied to determine the potential of the PBR for resisting nuclear-weapons proliferation and to optimize proliferation-resistant features. However, the purpose of this paper is to show that the method itself is viable. Refinements to the code are under way, with the objective of producing a powerful reactor physics

  2. Process for purifying graphite

    International Nuclear Information System (INIS)

    Clausius, R.A.

    1985-01-01

    A process for purifying graphite comprising: comminuting graphite containing mineral matter to liberate at least a portion of the graphite particles from the mineral matter; mixing the comminuted graphite particles containing mineral matter with water and hydrocarbon oil to form a fluid slurry; separating a water phase containing mineral matter and a hydrocarbon oil phase containing grahite particles; and separating the graphite particles from the hydrocarbon oil to obtain graphite particles reduced in mineral matter. Depending upon the purity of the graphite desired, steps of the process can be repeated one or more times to provide a progressively purer graphite

  3. Mechanics of binary and polydisperse spherical pebble assembly

    International Nuclear Information System (INIS)

    Annabattula, R.K.; Gan, Y.; Kamlah, M.

    2012-01-01

    The micromechanical behavior of an assembly of binary and polydisperse spherical pebbles is studied using discrete element method (DEM) accounting for microscopic interactions between individual pebbles. A in-house DEM code has been used to simulate the assemblies consisting of different pebble diameters and the results of the simulations are compared with that of mono-size pebble assemblies. The effect of relative radii and volume fraction of the pebbles on the macroscopic stress–strain response is discussed. Furthermore, the effect of packing factor and coefficient of friction on the overall stress–strain behavior of the system is studied in detail. The shear (tangential) stiffness between the particles is also another influencing parameter. For a very small shear stiffness the system shows a strong dependence on the packing factor while a pebble material dependent shear stiffness shows a rather moderate dependence on the packing factor. For a similar packing factor, the mono-size assembly shows a stiff behavior during loading compared to binary assembly. However, the simulations do not show a significant difference between the two behaviors in contrast to the observations made in the experiments. The discrepancy can be attributed to (i) probable difference in packing factors for mono-size and binary assemblies in the experiments, (ii) arbitrary friction coefficient in the current model and (iii) the tangential interaction (constant shear stiffness) implemented in the present model which needs further modification as a function of the load history on the pebbles. Evolution of other micromechanical characteristics such as coordination number, contact force distribution and stored elastic energy of individual pebbles as a function of external load and system parameters is presented which can be used to estimate important macroscopic properties such as overall thermal conductivity and crushing resistance of the pebble beds.

  4. Mechanics of binary and polydisperse spherical pebble assembly

    Energy Technology Data Exchange (ETDEWEB)

    Annabattula, R.K., E-mail: ratna.annabattula@kit.edu [Institute for Applied Materials (IAM-WBM), Karlsruhe Institute of Technology (KIT), D-76344, Eggenstein-Leopoldshafen (Germany); Gan, Y., E-mail: yixiang.gan@sydney.edu.au [School of Civil Engineering, University of Sydney, 2006 NSW, Sydney (Australia); Kamlah, M., E-mail: marc.kamlah@kit.edu [Institute for Applied Materials (IAM-WBM), Karlsruhe Institute of Technology (KIT), D-76344, Eggenstein-Leopoldshafen (Germany)

    2012-08-15

    The micromechanical behavior of an assembly of binary and polydisperse spherical pebbles is studied using discrete element method (DEM) accounting for microscopic interactions between individual pebbles. A in-house DEM code has been used to simulate the assemblies consisting of different pebble diameters and the results of the simulations are compared with that of mono-size pebble assemblies. The effect of relative radii and volume fraction of the pebbles on the macroscopic stress-strain response is discussed. Furthermore, the effect of packing factor and coefficient of friction on the overall stress-strain behavior of the system is studied in detail. The shear (tangential) stiffness between the particles is also another influencing parameter. For a very small shear stiffness the system shows a strong dependence on the packing factor while a pebble material dependent shear stiffness shows a rather moderate dependence on the packing factor. For a similar packing factor, the mono-size assembly shows a stiff behavior during loading compared to binary assembly. However, the simulations do not show a significant difference between the two behaviors in contrast to the observations made in the experiments. The discrepancy can be attributed to (i) probable difference in packing factors for mono-size and binary assemblies in the experiments, (ii) arbitrary friction coefficient in the current model and (iii) the tangential interaction (constant shear stiffness) implemented in the present model which needs further modification as a function of the load history on the pebbles. Evolution of other micromechanical characteristics such as coordination number, contact force distribution and stored elastic energy of individual pebbles as a function of external load and system parameters is presented which can be used to estimate important macroscopic properties such as overall thermal conductivity and crushing resistance of the pebble beds.

  5. The ESKOM pebble bed modular reactor

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1999-01-01

    An audit has been made of the design, construction, safety, economics and marketability of the ESKOM pebble bed modular reactor (PBMR). In this paper that audit is briefly summarized. The principal conclusions of the audit are as follows. The design is sound. It is a logical development of the designs proposed for other, modern, high-temperature gas-cooled reactors. More than 80% of the cost of constructing and commissioning a series of PBMRs would be spent in South Africa. The PBMR is much safer than existing nuclear power reactors and for many practical purposes it may be treated as a conventional chemical plant. The PBMR is economically competitive with thermal power stations. There is a substantial global market for the PBMR. (author)

  6. Failure initiation and propagation of Li4SiO4 pebbles in fusion blankets

    International Nuclear Information System (INIS)

    Zhao Shuo; Gan Yixiang; Kamlah, Marc

    2013-01-01

    Lithium orthosilicate (Li 4 SiO 4 ) pebbles are considered to be a candidate as solid tritium breeder in the helium cooled pebble bed (HCPB) blanket. These ceramic pebbles might be crushed during thermomechanical loading in the blanket. In this work, the failure initiation and propagation of pebbles in pebble beds is investigated using the discrete element method (DEM). Pebbles are simplified as mono-sized elastic spheres. Every pebble has a contact strength in terms of critical strain energy, which is derived from a validated strength model and crush test data for pebbles from a specific batch of Li 4 SiO 4 pebbles. Pebble beds are compressed uniaxially and triaxially in DEM simulations. When the strain energy absorbed by a pebble exceeds its critical energy it fails. The failure initiation is defined as a given small fraction of pebbles crushed. It is found that the load level for failure initiation can be very low. For example, if failure initiation is defined as soon as 0.02% of the pebbles have been crushed, the pressure required for uniaxial loading is about 2.5 MPa. Therefore, it is essential to study the influence of failure propagation on the macroscopic response of pebble beds. Thus a reduction ratio defined as the size ratio of a pebble before and after its failure is introduced. The macroscopic stress–strain relation is investigated with different reduction ratios. A typical stress plateau is found for a small reduction ratio.

  7. Estimation of graphite dust production in ITER TBM using finite element method

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Ji-Ho, E-mail: jhkang@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111, Daekeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Kim, Eung Seon [Korea Atomic Energy Research Institute, 989-111, Daekeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Ahn, Mu-Young; Lee, Youngmin; Park, Yi-Hyun; Cho, Seungyon [National Fusion Research Institute, 169-148, Gwahak-ro, Yuseong-gu, Daejeon (Korea, Republic of)

    2015-12-15

    Highlights: • Graphite dust production was estimated for the Korean Helium Cooled Ceramic Reflector. • Wear amount was calculated by Archard model using finite element analysis results. • Life time estimation of graphite dust production was done. - Abstract: In this study, an estimation method of graphite dust production in the pebble-bed type reflector region of the Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) of the International Thermonuclear Experimental Reactor (ITER) project using Finite Element Method (FEM) was proposed and the total amount of dust production was calculated. A unit-cell model of uniformly arranged pebbles was defined with thermal and mechanical loadings. A commercial FEM program, Abaqus V6.10, was used to model and solve the stress field under multiple contact constraints between pebbles in the unit-cell. Resultant normal contact forces and slip distances on the contact points were applied into the Archard adhesive wear model to calculate the amount of graphite dust. The Finite Element (FE) analysis was repeated at 27 unit-cell locations chosen to form an interpolated dust density function for the entire region of the reflector. The dust production calculation was extended to the life time of the HCCR and the total graphite dust production was estimated to 0.279 g at the end of the life time with the maximum graphite dust density of 0.149 μg/mm{sup 3}. The dust explosion could be a safety issue with the calculated dust density level and it requires that an appropriate maintenance to remove sufficient amount of graphite dust regularly to prevent the possibility of dust explosion.

  8. Estimation of graphite dust production in ITER TBM using finite element method

    International Nuclear Information System (INIS)

    Kang, Ji-Ho; Kim, Eung Seon; Ahn, Mu-Young; Lee, Youngmin; Park, Yi-Hyun; Cho, Seungyon

    2015-01-01

    Highlights: • Graphite dust production was estimated for the Korean Helium Cooled Ceramic Reflector. • Wear amount was calculated by Archard model using finite element analysis results. • Life time estimation of graphite dust production was done. - Abstract: In this study, an estimation method of graphite dust production in the pebble-bed type reflector region of the Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) of the International Thermonuclear Experimental Reactor (ITER) project using Finite Element Method (FEM) was proposed and the total amount of dust production was calculated. A unit-cell model of uniformly arranged pebbles was defined with thermal and mechanical loadings. A commercial FEM program, Abaqus V6.10, was used to model and solve the stress field under multiple contact constraints between pebbles in the unit-cell. Resultant normal contact forces and slip distances on the contact points were applied into the Archard adhesive wear model to calculate the amount of graphite dust. The Finite Element (FE) analysis was repeated at 27 unit-cell locations chosen to form an interpolated dust density function for the entire region of the reflector. The dust production calculation was extended to the life time of the HCCR and the total graphite dust production was estimated to 0.279 g at the end of the life time with the maximum graphite dust density of 0.149 μg/mm"3. The dust explosion could be a safety issue with the calculated dust density level and it requires that an appropriate maintenance to remove sufficient amount of graphite dust regularly to prevent the possibility of dust explosion.

  9. Thermo-mechanical characterization of ceramic pebbles for breeding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Lo Frano, Rosa, E-mail: rosa.lofrano@ing.unipi.it; Aquaro, Donato; Scaletti, Luca

    2016-11-01

    Highlights: • Experimental activities to characterize the Li{sub 4}SiO{sub 4}. • Compression tests of pebbles. • Experimental evaluation of thermal conductivity of pebbles bed at different temperatures. • Experimental test with/without compression load. - Abstract: An open issue for fusion power reactor is to design a suitable breeding blanket capable to produce the necessary quantity of the tritium and to transfer the energy of the nuclear fusion reaction to the coolant. The envisaged solution called Helium-Cooled Pebble Bed (HCPB) breeding blanket foresees the use of lithium orthosilicate (Li{sub 4}SiO{sub 4}) or lithium metatitanate (Li{sub 2}TiO{sub 3}) pebble beds. The thermal mechanical properties of the candidate pebble bed materials are presently extensively investigated because they are critical for the feasibility and performances of the numerous conceptual designs which use a solid breeder. This study is aimed at the investigation of mechanical properties of the lithium orthosilicate and at the characterization of the main chemical, physical and thermo-mechanical properties taking into account the production technology. In doing that at the Department of Civil and Industrial Engineering (DICI) of the University of Pisa adequate experiments were carried out. The obtained results may contribute to characterize the material of the pebbles and to optimize the design of the envisaged fusion breeding blankets.

  10. Vapour pressure of caesium over nuclear graphite

    International Nuclear Information System (INIS)

    Faircloth, R.L.; Pummery, F.C.W.

    1976-01-01

    The vapour pressure of caesium over a fine-grained isotropic moulded gilsocarbon nuclear graphite intended for use in the manufacture of fuel tubes for the high temperature reactor has been determined as a function of temperature and concentration by means of the Knudsen effusion technique. The concentration range 0 to 10 μg caesium/g graphite was investigated and it was concluded that a Langmuir adsorption situation exists under these conditions. (author)

  11. 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).

  12. A deceleration system for near-diameter spheres in pipeline transportation in a pebble bed reactor based on the resistance of a pneumatic cushion

    International Nuclear Information System (INIS)

    Liu, Hongbing; He, Ayada; Du, Dong; Wang, Xin; Zhang, Haiquan

    2014-01-01

    Highlights: • A deceleration system for fuel transportation in a pebble bed reactor is designed. • Dynamic analysis and motion analysis of the deceleration process are conducted. • The effectiveness of the system is verified by the analysis and the experiment. • Some key design parameters are studied to achieve effective deceleration. • This research provides a guide for the design of a pebble bed reactor. - Abstract: The fuel elements cycle occurring inside and outside the core of a pebble bed reactor is carried out by pneumatic conveying. In some processes of conveyance, it is necessary to reduce the velocity of the moving fuel element in a short time to avoid damage to the fuel elements and the equipment. In this research, a deceleration system for near-diameter spheres in pipeline transportation based on the resistance of a pneumatic cushion is designed to achieve an effective and reliable deceleration process. Dynamic analysis and motion analysis of the deceleration process are conducted. The results show that when the fuel element is moving in the deceleration pipeline, the gas in the pipeline is compressed to create a pneumatic cushion which resists the movement of the fuel element. In this way, the velocity of the fuel element is decreased to below the target value. During this process, the deceleration is steady and reliable. On this basis some key design parameters are studied, such as the deceleration pipeline length, the ratio of the diameter of the fuel element to the internal diameter of the pipeline, etc. The experimental results are generally consistent with the analysis and demonstrate the considerable effectiveness of the deceleration process as well. This research provides a guide for the design of the fuel elements cycling system in a pebble bed reactor along with the optimization of its control

  13. Determination of a geometry-dependent parameter and development of a calculation model for describing the fission products transport from spherical fuel elements of graphite moderated gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Weissfloch, R

    1973-07-15

    The fuel elements of high-temperature reactors, coated with pyrolitic carbon and covered with graphite, release fission products like all other fuel elements. Because of safety reasons, the rate of this release has to be kept low and has also to be predictable. Measured values from irradiation tests and from post-irradiation tests about the actual release of different fission products are presented. The physical and chemical mechanism, which determines the release, is extraordinarily complex and in particular not clearly defined. Because of the mentioned reasons, a simplified calculation model was developed, which only considers the release-mechanisms phenomenologically. This calculation model coincides very well in its results with values received in experiments until now. It can be held as an interim state on the way to a complete theory.

  14. Determination of a geometry-dependent parameter and development of a calculation model for describing the fission products transport from spherical fuel elements of graphite moderated gas-cooled reactors

    International Nuclear Information System (INIS)

    Weissfloch, R.

    The fuel elements of High-Temperature Reactors, coated with pyrolitic carbon and covered with graphite, release fission products like all other fuel elements. Because of safety reasons the rate of this release has to be kept low and has also to be predictable. Measured values from irradiation tests and from post-irradiation tests about the actual release of different fission products are present. The physical and chemical mechanism, which determines the release, is extraordinarily complex and in particular not clearly defined. Because of the mentioned reasons a simplified calculation model was developed, which only considers the release-mechanisms phenomenologically. This calculation model coincides very well in its results with values received in experiments until now. It can serve as an interim state on the way to a complete theory. (U.S.)

  15. Particle flow of ceramic breeder pebble beds in bi-axial compression experiments

    International Nuclear Information System (INIS)

    Hermsmeyer, S.; Reimann, J.

    2002-01-01

    Pebble beds of ceramic material are investigated within the framework of developing solid breeder blankets for future fusion power plants. A thermo-mechanical characterisation of such pebble beds is mandatory for understanding the behaviour of pebble beds, and thus the overall blanket, under fusion environment conditions. The mechanical behaviour of pebble beds is typically explored with uni-axial, bi-axial and tri-axial compression experiments. The latter two types of experiment are particularly revealing since they contain explicitly, beyond a compression behaviour of the bed, information on the conditions for pebble flow, i.e. macroscopic relocation, in the pebble bed. (orig.)

  16. 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.

  17. 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

  18. Pebble red modular reactor - South Africa

    International Nuclear Information System (INIS)

    Fox, M.; Mulder, E.

    1996-01-01

    In 1995 the South African Electricity Utility, ESKOM, was convinced of the economical advantages of high temperature gas-cooled reactors as viable supply side option. Subsequently planning of a techno/economic study for the year 1996 was initiated. Continuation to the construction phase of a prototype plant will depend entirely on the outcome of this study. A reactor plant of pebble bed design coupled with a direct helium cycle is perceived. The electrical output is limited to about 100 MW for reasons of safety, economics and flexibility. Design of the reactor will be based on internationally proven, available technology. An extended research and development program is not anticipated. New licensing rules and regulations will be required. Safety classification of components will be based on the merit of HTGR technology rather than attempting to adhere to traditional LWR rules. A medium term time schedule for the design and construction of a prototype plant, commissioning and performance testing is proposed during the years 2002 and 2003. Pending the performance outcome of this plant and the current power demand, series production of 100 MWe units is foreseen. (author)

  19. Identification of the key parameters defining the life of graphite core components

    International Nuclear Information System (INIS)

    Mitchell, M.N.

    2005-01-01

    The Core Structures of a Pebble Bed rector core comprise graphite reflectors constructed from blocks. These blocks are subject to high flux and temperatures as well as significant gradients in flux and temperature. This loading combined with the behaviour of graphite under irradiation gives rise to complex stress states within the reflector blocks. At some point, the stress state will reach a critical level and cracks will initiate within the blocks. The point of crack initiation is a useful point to define as the end of the part's life. The life of these graphite reflector parts in a pebble bed reactor (PBR) core determines the service life of the Core Structures. The replacement of the Core Structures' components will be a costly and time consuming. It is important that the components of the Core Structures be designed for the best life possible. As part of the conceptual design of the Pebble Bed Modular Reactor (PBMR), the assessment of the life of these components was examined. To facilitate the understanding of the parameters that influence the design life of the PBMR, a study has been completed into the effect of various design parameters on the design life of a typical side reflector block. Parameters investigated include: block geometry, material property variations, and load variations. The results of this study are to be presented. (author)

  20. Manufacturing Technology of Ceramic Pebbles for Breeding Blanket

    Directory of Open Access Journals (Sweden)

    Rosa Lo Frano

    2018-05-01

    Full Text Available An open issue for the fusion power reactor is the choice of breeding blanket material. The possible use of Helium-Cooled Pebble Breeder ceramic material in the form of pebble beds is of great interest worldwide as demonstrated by the numerous studies and research on this subject. Lithium orthosilicate (Li4SiO4 is a promising breeding material investigated in this present study because the neutron capture of Li-6 allows the production of tritium, 6Li (n, t 4He. Furthermore, lithium orthosilicate has the advantages of low activation characteristics, low thermal expansion coefficient, high thermal conductivity, high density and stability. Even if they are far from the industrial standard, a variety of industrial processes have been proposed for making orthosilicate pebbles with diameters of 0.1–1 mm. However, some manufacturing problems have been observed, such as in the chemical stability (agglomeration phenomena. The aim of this study is to provide a new methodology for the production of pebbles based on the drip casting method, which was jointly developed by the DICI-University of Pisa and Industrie Bitossi. Using this new (and alternative manufacturing technology, in the field of fusion reactors, appropriately sized ceramic pebbles could be produced for use as tritium breeders.

  1. Graphite Oxidation Thermodynamics/Reactions

    International Nuclear Information System (INIS)

    Propp, W.A.

    1998-01-01

    The vulnerability of graphite-matrix spent nuclear fuel to oxidation by the ambient atmosphere if the fuel canister is breached was evaluated. Thermochemical and kinetic data over the anticipated range of storage temperatures (200 to 400 C) were used to calculate the times required for a total carbon mass loss of 1 mgcm-2 from a fuel specimen. At 200 C, the time required to produce even this small loss is large, 900,000 yr. However, at 400 C the time required is only 1.9 yr. The rate of oxidation at 200 C is negligible, and the rate even at 400 C is so small as to be of no practical consequence. Therefore, oxidation of the spent nuclear fuel upon a loss of canister integrity is not anticipated to be a concern based upon the results of this study

  2. Effect of packing fraction variations on reactivity in pebble-bed reactor

    International Nuclear Information System (INIS)

    Snoj, L.; Ravnik, M.

    2004-01-01

    The pebble-bed reactor (PBR) core consists of large number of randomly packed spherical fuel elements. The effect of fuel element packing density variations on multiplication factor in a typical PBR is studied using WIMS code. It is observed that at normal conditions the k-eff increases with packing fraction. Effects of secondary coolant ingress (water or molten lead) in the core at accidental conditions are studied at various packing densities. The effect of water ingress on reactivity depends strongly on water density and packing fraction and is prevailingly positive, while the lead ingress reduces multiplication factor regardless of lead effective density and packing fraction. Both effects are stronger at lower packing fractions. (author)

  3. Letters initiating Clean Water Act 404(c) review of mining at Pebble deposit

    Science.gov (United States)

    Correspondence between EPA and the Pebble Limited Partnership and the State of Alaska initiating review under section 404(c) of the Clean Water Act of potential adverse environmental effects associated with mining the Pebble deposit in southwest Alaska.

  4. Phonon scattering in graphite

    International Nuclear Information System (INIS)

    Wagner, P.

    1976-04-01

    Effects on graphite thermal conductivities due to controlled alterations of the graphite structure by impurity addition, porosity, and neutron irradiation are shown to be consistent with the phonon-scattering formulation 1/l = Σ/sub i equals 1/sup/n/ 1/l/sub i/. Observed temperature effects on these doped and irradiated graphites are also explained by this mechanism

  5. Study on "1"4C content in post-irradiation graphite spheres of HTR-10

    International Nuclear Information System (INIS)

    Wang Shouang; Pi Yue; Xie Feng; Li Hong; Cao Jianzhu

    2014-01-01

    Since the production mechanism of the "1"4C in spherical fuel elements was similar to that of fuel-free graphite spheres, in order to obtain the amount of "1"4C in fuel elements and graphite spheres of HTR-10, the production mechanism of the "1"4C in graphite spheres was studied. The production sources of the "1"4C in graphite spheres and fuel elements were summarized, the amount of "1"4C in the post-irradiation graphite spheres was calculated, the decomposition techniques of graphite spheres were compared, and experimental methods for decomposing the graphite spheres and preparing the "1"4C sample were proposed. The results can lay the foundation for further experimental research and provide theoretical calculations for comparison. (authors)

  6. Conceptual study of ferromagnetic pebbles for heat exhaust in fusion reactors with short power decay length

    Directory of Open Access Journals (Sweden)

    N. Gierse

    2015-03-01

    The key results of this study are that very high heat fluxes are accessible in the operation space of ferromagnetic pebbles, that ferromagnetic pebbles are compatible with tokamak operation and current divertor designs, that the heat removal capability of ferromagnetic pebbles increases as λq decreases and, finally, that for fusion relevant values of q∥ pebble diameters below 100 μm are required.

  7. Tritium adsorption/release behaviour of advanced EU breeder pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Kolb, Matthias H.H., E-mail: matthias.kolb@kit.edu; Rolli, Rolf; Knitter, Regina

    2017-06-15

    The tritium loading of current grades of advanced ceramic breeder pebbles with three different lithium orthosilicate (LOS)/lithium metatitanate (LMT) compositions (20–30 mol% LMT in LOS) and pebbles of EU reference material, was performed in a consistent way. The temperature dependent release of the introduced tritium was subsequently investigated by temperature programmed desorption (TPD) experiments to gain insight into the desorption characteristics. The obtained TPD data was decomposed into individual release mechanisms according to well-established desorption kinetics. The analysis showed that the pebble composition of the tested samples does not severely change the release behaviour. Yet, an increased content of lithium metatitanate leads to additional desorption peaks at medium temperatures. The majority of tritium is released by high temperature release mechanisms of chemisorbed tritium, while the release of physisorbed tritium is marginal in comparison. The results allow valuable projections for the tritium release behaviour in a fusion blanket.

  8. Tritium adsorption/release behaviour of advanced EU breeder pebbles

    Science.gov (United States)

    Kolb, Matthias H. H.; Rolli, Rolf; Knitter, Regina

    2017-06-01

    The tritium loading of current grades of advanced ceramic breeder pebbles with three different lithium orthosilicate (LOS)/lithium metatitanate (LMT) compositions (20-30 mol% LMT in LOS) and pebbles of EU reference material, was performed in a consistent way. The temperature dependent release of the introduced tritium was subsequently investigated by temperature programmed desorption (TPD) experiments to gain insight into the desorption characteristics. The obtained TPD data was decomposed into individual release mechanisms according to well-established desorption kinetics. The analysis showed that the pebble composition of the tested samples does not severely change the release behaviour. Yet, an increased content of lithium metatitanate leads to additional desorption peaks at medium temperatures. The majority of tritium is released by high temperature release mechanisms of chemisorbed tritium, while the release of physisorbed tritium is marginal in comparison. The results allow valuable projections for the tritium release behaviour in a fusion blanket.

  9. Acceptance test for graphite components and construction status of HTTR

    International Nuclear Information System (INIS)

    Iyoku, T.; Ishihara, M.; Maruyama, S.; Shiozawa, S.; Tsuji, N.; Miki, T.

    1996-01-01

    In March, 1991, the Japan Atomic Energy Research Institute (JAERI) started to constructed the High Temperature engineering Test Reactor(HTTR) which is a 30-MW(thermal) helium gas-cooled reactor with a core composed of prismatic graphite blocks piled on the core support graphite structures. Two types of graphite materials are used in the HTTR. One is the garde IG-110, isotropic fine grain graphite, another is the grade PGX, medium-to-fine grained molded graphite. These materials were selected on the basis of the appropriate properties required by the HTTR reactor design. Industry-wide standards for an acceptance test of graphite materials used as main components of a nuclear reactor had not been established. The acceptance standard for graphite components of the HTTR, therefore, was drafted by JAERI and reviewed by specialists outside JAERI. The acceptance standard consists of the material testing, non-destructive examination such as the ultrasonic and eddy current testings, dimensional and visual inspections and assembly test. Ultrasonic and eddy current testings are applied to graphite logs to detect an internal flaw and to graphite components to detect a surface flaw, respectively. The assembly test is performed at the works, prior to their installation in the reactor pressure vessel, to examine fabricating precision of each component and alignment of piled-up structures. The graphite components of the HTTR had been tested on the basis of the acceptance standard. It was confirmed that the graphite manufacturing process was well controlled and high quality graphite components were provided to the HTTR. All graphite components except for the fuel graphite blocks are to be installed in the reactor pressure vessel of the HTTR in September 1995. The paper describes the construction status of the HTTR focusing on the graphite components. The acceptance test results are also presented in this paper. (author). Figs

  10. A virtual pebble game to ensemble average graph rigidity.

    Science.gov (United States)

    González, Luis C; Wang, Hui; Livesay, Dennis R; Jacobs, Donald J

    2015-01-01

    The body-bar Pebble Game (PG) algorithm is commonly used to calculate network rigidity properties in proteins and polymeric materials. To account for fluctuating interactions such as hydrogen bonds, an ensemble of constraint topologies are sampled, and average network properties are obtained by averaging PG characterizations. At a simpler level of sophistication, Maxwell constraint counting (MCC) provides a rigorous lower bound for the number of internal degrees of freedom (DOF) within a body-bar network, and it is commonly employed to test if a molecular structure is globally under-constrained or over-constrained. MCC is a mean field approximation (MFA) that ignores spatial fluctuations of distance constraints by replacing the actual molecular structure by an effective medium that has distance constraints globally distributed with perfect uniform density. The Virtual Pebble Game (VPG) algorithm is a MFA that retains spatial inhomogeneity in the density of constraints on all length scales. Network fluctuations due to distance constraints that may be present or absent based on binary random dynamic variables are suppressed by replacing all possible constraint topology realizations with the probabilities that distance constraints are present. The VPG algorithm is isomorphic to the PG algorithm, where integers for counting "pebbles" placed on vertices or edges in the PG map to real numbers representing the probability to find a pebble. In the VPG, edges are assigned pebble capacities, and pebble movements become a continuous flow of probability within the network. Comparisons between the VPG and average PG results over a test set of proteins and disordered lattices demonstrate the VPG quantitatively estimates the ensemble average PG results well. The VPG performs about 20% faster than one PG, and it provides a pragmatic alternative to averaging PG rigidity characteristics over an ensemble of constraint topologies. The utility of the VPG falls in between the most

  11. Graphite moderated 252Cf source

    International Nuclear Information System (INIS)

    Sajo B, L.; Barros, H.; Greaves, E. D.; Vega C, H. R.

    2014-08-01

    The thorium molten salt reactor is an attractive and affordable nuclear power option for developing countries with insufficient infrastructure and limited technological capability. In the aim of personnel training and experience gathering at the Universidad Simon Bolivar there is in progress a project of developing a subcritical thorium liquid fuel reactor. The neutron source to run this subcritical reactor is a 252 Cf source and the reactor will use high-purity graphite as moderator. Using the MCNP5 code the neutron spectra of the 252 Cf in the center of the graphite moderator has been estimated along the channel where the liquid thorium salt will be inserted; also the ambient dose equivalent due to the source has been determined around the moderator. (Author)

  12. Li2O-pebble type tritium breeding blanket for fusion experimental reactor, 1

    International Nuclear Information System (INIS)

    Tone, Tatsuzo; Iida, Hiromasa; Tanaka, Yoshihisa

    1984-01-01

    The fusion experimental reactor is the next stage device in Japan, which is planned to be constructed following the critical plasma experimental device JT-60 being constructed at present. The breeding blanket installed in nuclear fusion reactors is one of most important structures, and it is required to satisfy the fundamental performance of producing and continuously recovering tritium as the nuclear fusion fuel, and other requirement in good coordination. The Li 2 O pebble type breeding blanket that Kawasaki Heavy Industries Ltd. has examined is the concept for resolving the problems of the mass transfer and thermal stress cracking of Li 2 O, which are important in blanket design. In this paper, the concept and characteristics of this breeding blanket are discussed from the viewpoint of the breeding and continuous recovery of tritium, the ease of manufacture and the maintenance of soundness. The breeding blanket is composed of breeding region, tritium purge region, cooling region, plasma stabilizing conductors and blanket container. Li 2 O is excellent in its tritium breeding performance and heat conductivity. The functions required for the breeding blanket, the fundamental structure, the examples of breeding blanket concept, the selection of breeding blanket concept, the characteristics of Li 2 O pebble type blanket and its future prospect are described. (Kako, I.)

  13. Experimental Study and Computational Simulations of Key Pebble Bed Thermo-mechanics Issues for Design and Safety

    Energy Technology Data Exchange (ETDEWEB)

    Tokuhiro, Akira; Potirniche, Gabriel; Cogliati, Joshua; Ougouag, Abderrafi

    2014-07-08

    An experimental and computational study, consisting of modeling and simulation (M&S), of key thermal-mechanical issues affecting the design and safety of pebble-bed (PB) reactors was conducted. The objective was to broaden understanding and experimentally validate thermal-mechanic phenomena of nuclear grade graphite, specifically, spheres in frictional contact as anticipated in the bed under reactor relevant pressures and temperatures. The contact generates graphite dust particulates that can subsequently be transported into the flowing gaseous coolent. Under postulated depressurization transients and with the potential for leaked fission products to be adsorbed onto graphite 'dust', there is the potential for fission products to escape from the primary volume. This is a design safety concern. Furthermore, earlier safety assessment identified the distinct possibility for the dispersed dust to combust in contact with air if sufficient conditions are met. Both of these phenomena were noted as important to design review and containing uncertainty to warrant study. The team designed and conducted two separate effects tests to study and benchmark the potential dust-generation rate, as well as study the conditions under which a dust explosion may occure in a standardized, instrumented explosion chamber.

  14. Thermal conductivity of a graphite bipolar plate (BPP) and its thermal contact resistance with fuel cell gas diffusion layers: Effect of compression, PTFE, micro porous layer (MPL), BPP out-of-flatness and cyclic load

    Science.gov (United States)

    Sadeghifar, Hamidreza; Djilali, Ned; Bahrami, Majid

    2015-01-01

    This paper reports on measurements of thermal conductivity of a graphite bipolar plate (BPP) as a function of temperature and its thermal contact resistance (TCR) with treated and untreated gas diffusion layers (GDLs). The thermal conductivity of the BPP decreases with temperature and its thermal contact resistance with GDLs, which has been overlooked in the literature, is found to be dominant over a relatively wide range of compression. The effects of PTFE loading, micro porous layer (MPL), compression, and BPP out-of-flatness are also investigated experimentally. It is found that high PTFE loadings, MPL and even small BPP out-of-flatness increase the BPP-GDL thermal contact resistance dramatically. The paper also presents the effect of cyclic load on the total resistance of a GDL-BPP assembly, which sheds light on the behavior of these materials under operating conditions in polymer electrolyte membrane fuel cells.

  15. The investigation of HTGR fuel regeneration process

    Energy Technology Data Exchange (ETDEWEB)

    Lazarev, L N; Bertina, L E; Popik, V P; Isakov, V P; Alkhimov, N B; Pokhitonov, Yu A

    1985-07-01

    The aim of this report is the investigation of HTGR fuel regeneration. The operation in the technologic scheme of uranium extraction from fuel depleted elements is separation of fuel from graphite. Available methods of graphite matrix destruction are: mechanical destruction, chemical destruction, and burning. Mechanical destruction is done in combination with leaching or chlorination. Methods of chemical destruction of graphite matrix are not sufficiently studied. Most of the investigations nowadays sre devoted to removal of graphite by burning.

  16. The investigation of HTGR fuel regeneration process

    International Nuclear Information System (INIS)

    Lazarev, L.N.; Bertina, L.E.; Popik, V.P.; Isakov, V.P.; Alkhimov, N.B.; Pokhitonov, Yu.A.

    1985-01-01

    The aim of this report is the investigation of HTGR fuel regeneration. The operation in the technologic scheme of uranium extraction from fuel depleted elements is separation of fuel from graphite. Available methods of graphite matrix destruction are: mechanical destruction, chemical destruction, and burning. Mechanical destruction is done in combination with leaching or chlorination. Methods of chemical destruction of graphite matrix are not sufficiently studied. Most of the investigations nowadays sre devoted to removal of graphite by burning

  17. Graphite moderated reactor for thermoelectric generation

    International Nuclear Information System (INIS)

    Akazawa, Issei; Yamada, Akira; Mizogami, Yorikata

    1998-01-01

    Fuel rods filled with cladded fuel particles distributed and filled are buried each at a predetermined distance in graphite blocks situated in a reactor core. Perforation channels for helium gas as coolants are formed to the periphery thereof passing through vertically. An alkali metal thermoelectric power generation module is disposed to the upper lid of a reactor container while being supported by a securing receptacle. Helium gas in the coolant channels in the graphite blocks in the reactor core absorbs nuclear reaction heat, to be heated to a high temperature, rises upwardly by the reduction of the specific gravity, and then flows into an upper space above the laminated graphite block layer. Then the gas collides against a ceiling and turns, and flows down in a circular gap around the circumference of the alkali metal thermoelectric generation module. In this case, it transfers heat to the alkali metal thermoelectric generation module. (I.N.)

  18. The modular pebble bed nuclear reactor - the preferred new sustainable energy source for electricity, hydrogen and potable water production?

    International Nuclear Information System (INIS)

    Kemeny, L.G.

    2003-01-01

    This paper describes a joint project of Massachusetts Institute of technology, Nu-Tec Inc. and Proto Power. The elegant simplicity of graphite moderated pebble bed reactor is the basis for the 'generation four' nuclear power plants. High Temperature Gas Cooled (HTGC) nuclear power plant have the potential to become the preferred base load sustainable energy source for the new millennium. The great attraction of these helium cooled 'Generation Four' nuclear plant can be summarised as follows: Factory assembly line production; Modularity and ease of delivery to site; High temperature Brayton Cycle ideally suited for cogeneration of electricity, potable water and hydrogen; Capital and operating costs competitive with hydrocarbon plant; Design is inherently meltdown proof and proliferation resistant

  19. Uraniferous quartz-pebble conglomerates in South Africa

    International Nuclear Information System (INIS)

    von Backstroem, J.W.

    1981-01-01

    The purpose of this paper is to give a short background statement summarizing data on the Dominion Reef Group, the Witwatersrand Supergroup, and the Ventersdorp Contact Reef, with particular reference to the close relationship of gold and uranium with sedimentary features as well as the mineralization, conditions of deposition, and the nature of the quartz-pebble conglomerates

  20. TEM study of impurity segregations in beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Klimenkov, M., E-mail: michael.klimenkov@kit.edu [Institute for Applied Materials – Applied Materials Physics, Karlsruhe Institute of Technology, Hermann-von-Helmholz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Chakin, V.; Moeslang, A. [Institute for Applied Materials – Applied Materials Physics, Karlsruhe Institute of Technology, Hermann-von-Helmholz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rolli, R. [Institute for Applied Materials – Materials and Biomechanics, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2014-12-15

    Beryllium is planned to be used as a neutron multiplier in the Helium-cooled Pebble Bed European concept of a breeding blanket of demonstration power reactor DEMO. In order to evaluate the irradiation performance, individual pebbles and constrained pebble beds were neutron-irradiated at temperatures typical of fusion blankets. Beryllium pebbles 1 mm in diameter produced by the rotating electrode method were subjected to a TEM study before and after irradiation at High Flux Reactor, Petten, Netherlands at 861 K. The grain size varied in a wide range from sub-micron size up to several tens of micrometers, which indicated formation bimodal grain size distribution. Based on the application of combined electron energy loss spectroscopy and energy dispersive X-ray spectroscopy methods, we suggest that impurity precipitates play an important role in controlling the mechanical properties of beryllium. The impurity elements were present in beryllium at a sub-percent concentration form beryllide particles of a complex (Fe/Al/Mn/Cr)B composition. These particles are often ordered along dislocations lines, forming several micron-long chains. It can be suggested that fracture surfaces often extended along these chains in irradiated material.

  1. Transmutation of plutonium in pebble bed type high temperature reactors

    International Nuclear Information System (INIS)

    Bende, E.E.

    1997-01-01

    The pebble bed type High Temperature Reactor (HTR) has been studied as a uranium-free burner of reactor grade plutonium. In a parametric study, the plutonium loading per pebble as well as the type and size of the coated particles (CPs) have been varied to determine the plutonium consumption, the final plutonium burnup, the k ∞ and the temperature coefficients as a function of burnup. The plutonium loading per pebble is bounded between 1 and 3 gr Pu per pebble. The upper limit is imposed by the maximal allowable fast fluence for the CPs. A higher plutonium loading requires a longer irradiation time to reach a desired burnup, so that the CPs are exposed to a higher fast fluence. The lower limit is determined by the temperature coefficients, which become less negative with increasing moderator-actinide ratio. A burnup of about 600 MWd/kgHM can be reached. With the HTR's high efficiency of 40%, a plutonium supply of 1520 kg/GW e a is achieved. The discharges of plutonium and minor actinides are then 450 and 110 kg/GW e a, respectively. (author)

  2. Performance Evaluation of a Pebble Bed Solar Crop Dryer ...

    African Journals Online (AJOL)

    Nigerian Journal of Technology ... The solar crop dryer consists of an imbedded pebble bed solar heat storage unit/solar collector ... The crop-drying chamber is made of drying trays of wire gauze while the roof is made of transparent glazing.

  3. Fabrication of Li2TiO3 pebbles by a freeze drying process

    International Nuclear Information System (INIS)

    Lee, Sang-Jin; Park, Yi-Hyun; Yu, Min-Woo

    2013-01-01

    Li 2 TiO 3 pebbles were successfully fabricated by using a freeze drying process. The Li 2 TiO 3 slurry was prepared using a commercial powder of particle size 0.5–1.5 μm and the pebble pre-form was prepared by dropping the slurry into liquid nitrogen through a syringe needle. The droplets were rapidly frozen, changing their morphology to spherical pebbles. The frozen pebbles were dried at −10 °C in vacuum. To make crack-free pebbles, some glycerin was employed in the slurry, and long drying time and a low vacuum condition were applied in the freeze drying process. In the process, the solid content in the slurry influenced the spheroidicity of the pebble green body. The dried pebbles were sintered at 1200 °C in an air atmosphere. The sintered pebbles showed almost 40% shrinkage. The sintered pebbles revealed a porous microstructure with a uniform pore distribution and the sintered pebbles were crushed under an average load of 50 N in a compressive strength test. In the present study, a freeze drying process for fabrication of spherical Li 2 TiO 3 pebbles is introduced. The processing parameters, such as solid content in the slurry and the conditions of freeze drying and sintering, are also examined

  4. Pore Scale Thermal Hydraulics Investigations of Molten Salt Cooled Pebble Bed High Temperature Reactor with BCC and FCC Configurations

    Directory of Open Access Journals (Sweden)

    Shixiong Song

    2014-01-01

    CFD results and empirical correlations’ predictions of pressure drop and local Nusselt numbers. Local pebble surface temperature distributions in several default conditions are investigated. Thermal removal capacities of molten salt are confirmed in the case of nominal condition; the pebble surface temperature under the condition of local power distortion shows the tolerance of pebble in extreme neutron dose exposure. The numerical experiments of local pebble insufficient cooling indicate that in the molten salt cooled pebble bed reactor, the pebble surface temperature is not very sensitive to loss of partial coolant. The methods and results of this paper would be useful for optimum designs and safety analysis of molten salt cooled pebble bed reactors.

  5. Modelling of thermal and mechanical behaviour of pebble beds

    International Nuclear Information System (INIS)

    Boccaccini, L.V.; Buehler, L.; Hermsmeyer, S.; Wolf, F.

    2001-01-01

    FZK (Forshungzentrum Karlsruhe) is developing a Helium Cooled Pebble Bed (HCPB) Blanket Concept for fusion power reactors based on the use of ceramic breeder materials and beryllium multiplier in the form of pebble beds. The design of such a blanket requires models and computer codes describing the thermal-mechanical behavior of pebble beds to evaluate the temperatures, stresses, deformations and mechanical interactions between pebble beds and the structure with required accuracy and reliability. The objective to describe the beginning of life condition for the HCPB blanket seems near to be reached. Mechanical models that describe the thermo-mechanical behavior of granular materials used in form of pebble beds are implemented in a commercial structure code. These models have been calibrated using the results of a large series of dedicated experiments. The modeling work is practically concluded for ceramic breeder; it will be carried on in the next year for beryllium to obtain the required correlations for creep and the thermal conductivity. The difficulties for application in large components (such as the HCPB blanket) are the limitations of the present commercial codes to manage such a set of constitutive equations under complex load conditions and large mesh number. The further objective is to model the thermal cycles during operation; the present correlations have to be adapted for the release phase. A complete description of the blanket behavior during irradiation is at the present out of our capability; this objective requires an extensive R and D program that at the present is only at the beginning. (Y.Tanaka)

  6. Experimental and numerical validation of a two-region-designed pebble bed reactor with dynamic core

    International Nuclear Information System (INIS)

    Jiang, S.Y.; Yang, X.T.; Tang, Z.W.; Wang, W.J.; Tu, J.Y.; Liu, Z.Y.; Li, J.

    2012-01-01

    Highlights: ► The experimental installation has been built to investigate the pebble flow. ► The feasibility of two-region pebble bed reactor has been verified. ► The pebble flow is more uniform in a taller vessel than that in a lower vessel. ► Larger base cone angle will decrease the scale of the stagnant zone. - Abstract: The pebble flow is the principal issue for the design of the pebble bed reactor. In order to verify the feasibility of a two-region-designed pebble bed reactor, the experimental installation with a taller vessel has been built, which is proportional to the real pebble bed reactor. With the aid of the experimental installation, the stable establishment and maintenance of the two-region arrangement has been verified, at the same time, the applicability of the DEM program has been also validated. Research results show: (1) The pebble's bouncing on the free surface is an important factor for the mixing of the different colored pebbles. (2) Through the guide plates installed in the top of the pebble packing, the size of the mixing zone can be reduced from 6–7 times to 3–4 times the pebble diameter. (3) The relationship between the width of the central region and the ratio of loading pebbles is approximately linear in the taller vessel. (4) The heighten part of the pebble packing can improve the uniformity of the flowing in the lower. (5) To increase the base cone angle can decrease the scale of the stagnant zone. All of these conclusions are meaningful to the design of the real pebble reactor.

  7. Hydrogen storage in graphitic nanofibres

    OpenAIRE

    McCaldin, Simon Roger

    2007-01-01

    There is huge need to develop an alternative to hydrocarbons fuel, which does not produce CO2 or contribute to global warming - 'the hydrogen economy' is such an alternative, however the storage of hydrogen is the key technical barrier that must be overcome. The potential of graphitic nanofibres (GNFs) to be used as materials to allow the solid-state storage of hydrogen has thus been investigated. This has been conducted with a view to further developing the understanding of the mechanism(s) ...

  8. Progress in radioactive graphite waste management. Additional information

    International Nuclear Information System (INIS)

    2010-06-01

    Radioactive graphite constitutes a major waste stream which arises during the decommissioning of certain types of nuclear installations. Worldwide, a total of around 250 000 tonnes of radioactive graphite, comprising graphite moderators and reflectors, will require management solutions in the coming years. 14 C is the radionuclide of greatest concern in nuclear graphite; it arises principally through the interaction of reactor neutrons with nitrogen, which is present in graphite as an impurity or in the reactor coolant or cover gas. 3 H is created by the reactions of neutrons with 6 Li impurities in graphite as well as in fission of the fuel. 36 Cl is generated in the neutron activation of chlorine impurities in graphite. Problems in the radioactive waste management of graphite arise mainly because of the large volumes requiring disposal, the long half-lives of the main radionuclides involved and the specific properties of graphite - such as stored Wigner energy, graphite dust explosibility and the potential for radioactive gases to be released. Various options for the management of radioactive graphite have been studied but a generally accepted approach for its conditioning and disposal does not yet exist. Different solutions may be appropriate in different cases. In most of the countries with radioactive graphite to manage, little progress has been made to date in respect of the disposal of this material. Only in France has there been specific thinking about a dedicated graphite waste-disposal facility (within ANDRA): other major producers of graphite waste (UK and the countries of the former Soviet Union) are either thinking in terms of repository disposal or have no developed plans. A conference entitled 'Solutions for Graphite Waste: a Contribution to the Accelerated Decommissioning of Graphite Moderated Nuclear Reactors' was held at the University of Manchester 21-23 March 2007 in order to stimulate progress in radioactive graphite waste management

  9. A graphite nanoeraser

    DEFF Research Database (Denmark)

    Liu, Ze; Bøggild, Peter; Yang, Jia-rui

    2011-01-01

    We present here a method for cleaning intermediate-size (up to 50 nm) contamination from highly oriented pyrolytic graphite and graphene. Electron-beam-induced deposition of carbonaceous material on graphene and graphite surfaces inside a scanning electron microscope, which is difficult to remove...... by conventional techniques, can be removed by direct mechanical wiping using a graphite nanoeraser, thus drastically reducing the amount of contamination. We discuss potential applications of this cleaning procedure....

  10. Gas reactor international cooperative program interim report: German Pebble Bed Reactor design and technology review

    International Nuclear Information System (INIS)

    1978-09-01

    This report describes and evaluates several gas-cooled reactor plant concepts under development within the Federal Republic of Germany (FRG). The concepts, based upon the use of a proven Pebble Bed Reactor (PBR) fuel element design, include nuclear heat generation for chemical processes and electrical power generation. Processes under consideration for the nuclear process heat plant (PNP) include hydrogasification of coal, steam gasification of coal, combined process, and long-distance chemical heat transportation. The electric plant emphasized in the report is the steam turbine cycle (HTR-K), although the gas turbine cycle (HHT) is also discussed. The study is a detailed description and evaluation of the nuclear portion of the various plants. The general conclusions are that the PBR technology is sound and that the HTR-K and PNP plant concepts appear to be achievable through appropriate continuing development programs, most of which are either under way or planned

  11. Fundamental burn-up mode in a pebble-bed type reactor

    International Nuclear Information System (INIS)

    Chen, Xue-Nong; Kiefhaber, Edgar; Maschek, Werner

    2008-01-01

    This paper deals with a pebble-bed type reactor, in which the fuel is loaded from one side (top) and discharged from the other side (bottom). A boundary value problem of a single group diffusion equation coupled with simplified burn-up equations is studied, where the natural radioactive decay processes are neglected in the burn-up modelling. An asymptotic burning wave solution is found analytically in the one-dimensional case, which is called as fundamental burn-up mode. Among this solution family there are two particular cases, namely, a classic fundamental solution with a zero burn-up and a partial solitary burn-up wave solution with a highest burn-up. An example of Th-U conversion is considered and the solutions are presented in order to show the mechanism of the burning wave. (author)

  12. Oxidation Resistant Graphite Studies

    Energy Technology Data Exchange (ETDEWEB)

    W. Windes; R. Smith

    2014-07-01

    The Very High Temperature Reactor (VHTR) Graphite Research and Development Program is investigating doped nuclear graphite grades exhibiting oxidation resistance. During a oxygen ingress accident the oxidation rates of the high temperature graphite core region would be extremely high resulting in significant structural damage to the core. Reducing the oxidation rate of the graphite core material would reduce the structural effects and keep the core integrity intact during any air-ingress accident. Oxidation testing of graphite doped with oxidation resistant material is being conducted to determine the extent of oxidation rate reduction. Nuclear grade graphite doped with varying levels of Boron-Carbide (B4C) was oxidized in air at nominal 740°C at 10/90% (air/He) and 100% air. The oxidation rates of the boronated and unboronated graphite grade were compared. With increasing boron-carbide content (up to 6 vol%) the oxidation rate was observed to have a 20 fold reduction from unboronated graphite. Visual inspection and uniformity of oxidation across the surface of the specimens were conducted. Future work to determine the remaining mechanical strength as well as graphite grades with SiC doped material are discussed.

  13. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 5: Graphite PIRTs

    International Nuclear Information System (INIS)

    Burchell, Timothy D.; Bratton, Rob; Marsden, Barry; Srinivasan, Makuteswara; Penfield, Scott; Mitchell, Mark; Windes, Will

    2008-01-01

    Here we report the outcome of the application of the Nuclear Regulatory Commission (NRC) Phenomena Identification and Ranking Table (PIRT) process to the issue of nuclear-grade graphite for the moderator and structural components of a next generation nuclear plant (NGNP), considering both routine (normal operation) and postulated accident conditions for the NGNP. The NGNP is assumed to be a modular high-temperature gas-cooled reactor (HTGR), either a gas-turbine modular helium reactor (GTMHR) version (a prismatic-core modular reactor (PMR)] or a pebble-bed modular reactor (PBMR) version (a pebble bed reactor (PBR)] design, with either a direct- or indirect-cycle gas turbine (Brayton cycle) system for electric power production, and an indirect-cycle component for hydrogen production. NGNP design options with a high-pressure steam generator (Rankine cycle) in the primary loop are not considered in this PIRT. This graphite PIRT was conducted in parallel with four other NRC PIRT activities, taking advantage of the relationships and overlaps in subject matter. The graphite PIRT panel identified numerous phenomena, five of which were ranked high importance-low knowledge. A further nine were ranked with high importance and medium knowledge rank. Two phenomena were ranked with medium importance and low knowledge, and a further 14 were ranked medium importance and medium knowledge rank. The last 12 phenomena were ranked with low importance and high knowledge rank (or similar combinations suggesting they have low priority). The ranking/scoring rationale for the reported graphite phenomena is discussed. Much has been learned about the behavior of graphite in reactor environments in the 60-plus years since the first graphite rectors went into service. The extensive list of references in the Bibliography is plainly testament to this fact. Our current knowledge base is well developed. Although data are lacking for the specific grades being considered for Generation IV (Gen IV

  14. Contribution to the study of can deformations in the fuel elements of gas-graphite reactors during thermal cycling; Contribution a l'etude des deformations des gaines des elements combustibles de reacteur graphite-gaz au cours du cyclage thermique

    Energy Technology Data Exchange (ETDEWEB)

    Gauthron, M; Boudouresques, B; Delpeyroux, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The cans of fuel cartridges used in reactors of the gas-graphite type have either longitudinal fins of variable thickness, short herring-bone fins, or else a mixture of the two. An important test of the strength of these cartridges is their behaviour during thermal cycling carried out in cells reproducing in-pile conditions. It has been observed during with rapid cooling that there occurs a shortening at the base of the fins which can be accompanied in particular by a compression effect at the fin type, which has a tendency to curl, and by a tractive force acting on the body of the can at the ends of the longitudinal fins; this last phenomenon can result in a fracturing of the welds at the extremities or of the ends of the cartridge. This report presents first of all the way in which the stress diagram can be drawn for a can touching the fuel, and then the effect of the ratchet along a fin fixed to a bar with or without grooves. Finally the importance is shown of the test cycling variables (temperature, heating and cooling rates). (authors) [French] Les gaines des cartouches combustibles des reacteurs de la filiere graphite-gaz comportent soit des ailettes longitudinales plus ou moins epaisses, soit de courtes ailettes a chevrons, soit un ensemble des deux. Un test important de la tenue des cartouches, est la tenue au cyclage thermique en cellule pour reproduire le comportement en pile. On a observe au cours des cyclages a refroidissement rapide, un raccourcissement a la base des ailettes qui peut s'accompagner notamment d'une mise en compression du sommet de l'ailette qui a tendance a friser, et d'une traction exercee sur le corps des gaines au bout des ailettes longitudinales; ce dernier phenomene peut se traduire par des ruptures de soudures d'extremites ou des parties terminales de la cartouche. Ce rapport presente d'abord la maniere dont peut etre trace le diagramme des contraintes dans une gaine liee au combustible, puis l'effet du rochet le long d

  15. Method for producing dustless graphite spheres from waste graphite fines

    Science.gov (United States)

    Pappano, Peter J [Oak Ridge, TN; Rogers, Michael R [Clinton, TN

    2012-05-08

    A method for producing graphite spheres from graphite fines by charging a quantity of spherical media into a rotatable cylindrical overcoater, charging a quantity of graphite fines into the overcoater thereby forming a first mixture of spherical media and graphite fines, rotating the overcoater at a speed such that the first mixture climbs the wall of the overcoater before rolling back down to the bottom thereby forming a second mixture of spherical media, graphite fines, and graphite spheres, removing the second mixture from the overcoater, sieving the second mixture to separate graphite spheres, charging the first mixture back into the overcoater, charging an additional quantity of graphite fines into the overcoater, adjusting processing parameters like overcoater dimensions, graphite fines charge, overcoater rotation speed, overcoater angle of rotation, and overcoater time of rotation, before repeating the steps until graphite fines are converted to graphite spheres.

  16. Graphite targets at LAMPF

    International Nuclear Information System (INIS)

    Brown, R.D.; Grisham, D.L.

    1983-01-01

    Rotating polycrystalline and stationary pyrolytic graphite target designs for the LAMPF experimental area are described. Examples of finite element calculations of temperatures and stresses are presented. Some results of a metallographic investigation of irradiated pyrolytic graphite target plates are included, together with a brief description of high temperature bearings for the rotating targets

  17. Electrochemical treatment of graphite

    Energy Technology Data Exchange (ETDEWEB)

    Podlovilin, V.I.; Egorov, I.M.; Zhernovoj, A.I.

    1983-01-01

    In the course of investigating various modes of electrochemical treatment (ECT) it has been found that graphite anode treatment begins under the ''glow mode''. A behaviour of some marks of graphite with the purpose of ECT technique development in different electrolytes has been tested. Electrolytes have been chosen of three types: highly alkaline (pH 13-14), neutral (pH-Z) and highly acidic (pH 1-2). For the first time parallel to mechanical electroerosion treatment, ECT of graphite and carbon graphite materials previously considered chemically neutral is proposed. ECT of carbon graphite materials has a number of advantages as compared with electroerrosion and mechanical ones with respect to the treatment rate and purity (ronghness) of the surface. A small quantity of sludge (6-8%) under ECT is in highly alkali electrolytes.

  18. Electrochemical treatment of graphite

    International Nuclear Information System (INIS)

    Podlovilin, V.I.; Egorov, I.M.; Zhernovoj, A.I.

    1983-01-01

    In the course of investigating various modes of electroche-- mical treatment (ECT) it has been found that graphite anode treatment begins under the ''glow mode''. A behaviour of some marks of graphite with the purpose of ECT technique development in different electrolytes has been tested. Electrolytes have been chosen of three types: highly alkaline (pH 13-14), neutral (pH-Z) and highly acidic (pH 1-2). For the first time parallel to mechanical electroerosion treatment ECT graphite and carbon graphite materials previously considered chemically neutral is proposed. ECT of carbon graphite materials has a number of advantages as compared with electroerrosion and mechanical ones this is treatment rate and purity (ronghness) of the surface. A sMall quantity of sludge (6-8%) under ECT is in highly alkali electrolytes

  19. Development of integrated waste management options for irradiated graphite

    Directory of Open Access Journals (Sweden)

    Alan Wareing

    2017-08-01

    Full Text Available The European Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project sought to develop best practices in the retrieval, treatment, and disposal of irradiated graphite including other irradiated carbonaceous waste such as structural material made of graphite, nongraphitized carbon bricks, and fuel coatings. Emphasis was given on legacy irradiated graphite, as this represents a significant inventory in respective national waste management programs. This paper provides an overview of the characteristics of graphite irradiated during its use, primarily as a moderator material, within nuclear reactors. It describes the potential techniques applicable to the retrieval, treatment, recycling/reuse, and disposal of these graphite wastes. Considering the lifecycle of nuclear graphite, from manufacture to final disposal, a number of waste management options have been developed. These options consider the techniques and technologies required to address each stage of the lifecycle, such as segregation, treatment, recycle, and ultimate disposal in a radioactive waste repository, providing a toolbox to aid operators and regulators to determine the most appropriate management strategy. It is noted that national waste management programs currently have, or are in the process of developing, respective approaches to irradiated graphite management. The output of the Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project is intended to aid these considerations, rather than dictate them.

  20. Characteristics of first loaded IG-110 graphite in HTTR core

    International Nuclear Information System (INIS)

    Sumita, Junya; Shibata, Taiju; Iyoku, Tatsuo; Sawa, Kazuhiro; Hanawa, Satoshi; Ishihara, Masahiro

    2006-10-01

    IG-110 graphite is a fine-grained isotropic and nuclear-grade graphite with excellent resistivity on both irradiation and corrosion and with high strength. The IG-110 graphite is used for the graphite components of High Temperature Engineering Test Reactor (HTTR) such as fuel and control rod guide blocks and support posts. In order to design and fabricate the graphite components in the HTTR, the Japan Atomic Energy Research Institute (the Japan Atomic Energy Agency at present) had established the graphite structural design code and design data on the basis of former research results. After the design code establishment, the IG-110 graphite components were fabricated and loaded in the HTTR core. This report summarized the characteristics of the first loaded IG-110 graphite as basic data for surveillance test, measuring material characteristics changed by neutron irradiation and oxidation. By comparing the design data, it was shown that the first loaded IG-110 graphite had excellent strength properties and enough safety margins to the stress limits in the design code. (author)

  1. Development of integrated waste management options for irradiated graphite

    Energy Technology Data Exchange (ETDEWEB)

    Wareing, Alan; Abrahamsen-Mills, Liam; Fowler, Linda; Jarvis, Richard; Banford, Anthony William [National Nuclear Laboratory, Warrington (United Kingdom); Grave, Michael [Doosan Babcock, Gateshead (United Kingdom); Metcalfe, Martin [National Nuclear Laboratory, Gloucestershire (United Kingdom); Norris, Simon [Radioactive Waste Management Limited, Oxon (United Kingdom)

    2017-08-15

    The European Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project sought to develop best practices in the retrieval, treatment, and disposal of irradiated graphite including other irradiated carbonaceous waste such as structural material made of graphite, nongraphitized carbon bricks, and fuel coatings. Emphasis was given on legacy irradiated graphite, as this represents a significant inventory in respective national waste management programs. This paper provides an overview of the characteristics of graphite irradiated during its use, primarily as a moderator material, within nuclear reactors. It describes the potential techniques applicable to the retrieval, treatment, recycling/reuse, and disposal of these graphite wastes. Considering the lifecycle of nuclear graphite, from manufacture to final disposal, a number of waste management options have been developed. These options consider the techniques and technologies required to address each stage of the lifecycle, such as segregation, treatment, recycle, and ultimate disposal in a radioactive waste repository, providing a toolbox to aid operators and regulators to determine the most appropriate management strategy. It is noted that national waste management programs currently have, or are in the process of developing, respective approaches to irradiated graphite management. The output of the Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project is intended to aid these considerations, rather than dictate them.

  2. Uranium assessment for the Precambrian pebble conglomerates in southeastern Wyoming

    International Nuclear Information System (INIS)

    Borgman, L.E.; Sever, C.; Quimby, W.F.; Andrew, M.E.; Karlstrom, K.E.; Houston, R.S.

    1981-03-01

    This volume is a geostatistical resource estimate of uranium and thorium in quartz-pebble conglomerates, and is a companion to Volume 1: The Geology and Uranium Potential to Precambrian Conglomerates in the Medicine Bow Mountains and Sierra Madre of Southeastern Wyoming; and to Volume 2: Drill-Hole Data, Drill-Site Geology, and Geochemical Data from the Study of Precambrian Uraniferous Conglomerates of the Medicine Bow Mountains and the Sierra Madre of Southeastern Wyoming

  3. Thermal cycling tests on Li4SiO4 and beryllium pebbles

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Norajitra, P.; Weisenburger, A.

    1995-01-01

    The European B.O.T. Demo-relevant solid breeder blanket is based on the use of beds of beryllium and Li 4 SiO 4 pebbles. Particularly dangerous for the pebble integrity are the rapid temperature changes which could occur, for instance, by a sudden blanket power shut-down. A series of thermal cycle tests have been performed for various beds of beryllium and Li 4 SiO 4 pebbles. No breaking was observed in the beryllium pebbles, however the Li 4 SiO 4 pebbles broke by temperature rates of change of about -50 C/sec independently on pebbles size and lithium enrichment. This value is considerably higher than the peak temperature rates of change expected in the blanket. (orig.)

  4. Asymptomatic Intracorneal Graphite Deposits following Graphite Pencil Injury

    OpenAIRE

    Philip, Swetha Sara; John, Deepa; John, Sheeja Susan

    2012-01-01

    Reports of graphite pencil lead injuries to the eye are rare. Although graphite is considered to remain inert in the eye, it has been known to cause severe inflammation and damage to ocular structures. We report a case of a 12-year-old girl with intracorneal graphite foreign bodies following a graphite pencil injury.

  5. TRISO Fuel Performance: Modeling, Integration into Mainstream Design Studies, and Application to a Thorium-fueled Fusion-Fission Hybrid Blanket

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey James [Univ. of California, Berkeley, CA (United States)

    2011-11-30

    This study focused on creating a new tristructural isotropic (TRISO) coated particle fuel performance model and demonstrating the integration of this model into an existing system of neutronics and heat transfer codes, creating a user-friendly option for including fuel performance analysis within system design optimization and system-level trade-off studies. The end product enables both a deeper understanding and better overall system performance of nuclear energy systems limited or greatly impacted by TRISO fuel performance. A thorium-fueled hybrid fusion-fission Laser Inertial Fusion Energy (LIFE) blanket design was used for illustrating the application of this new capability and demonstrated both the importance of integrating fuel performance calculations into mainstream design studies and the impact that this new integrated analysis had on system-level design decisions. A new TRISO fuel performance model named TRIUNE was developed and verified and validated during this work with a novel methodology established for simulating the actual lifetime of a TRISO particle during repeated passes through a pebble bed. In addition, integrated self-consistent calculations were performed for neutronics depletion analysis, heat transfer calculations, and then fuel performance modeling for a full parametric study that encompassed over 80 different design options that went through all three phases of analysis. Lastly, side studies were performed that included a comparison of thorium and depleted uranium (DU) LIFE blankets as well as some uncertainty quantification work to help guide future experimental work by assessing what material properties in TRISO fuel performance modeling are most in need of improvement. A recommended thorium-fueled hybrid LIFE engine design was identified with an initial fuel load of 20MT of thorium, 15% TRISO packing within the graphite fuel pebbles, and a 20cm neutron multiplier layer with beryllium pebbles in flibe molten salt coolant. It operated

  6. Development of Chinese HTR-PM pebble bed equivalent conductivity test facility

    Energy Technology Data Exchange (ETDEWEB)

    Ren, Cheng; Yang, Xingtuan; Jiang, Shengyao [Tsinghua Univ., Beijing (China). Inst. of Nuclear and New Energy Technology

    2016-01-15

    The first two 250-MWt high-temperature reactor pebble bed modules (HTR-PM) have been installing at the Shidaowan plant in Shandong Province, China. The values of the effective thermal conductivity of the pebble bed core are essential parameters for the design. For their determination, Tsinghua University in China has proposed a full-scale heat transfer experiment to conduct comprehensive thermal transfer tests in packed pebble bed and to determine the effective thermal conductivity.

  7. Numerical investigation of the flow at the pebble bed of the high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    Costa, Franklin C.; Navarro, Moyses A.; Santos, Andre A.C.

    2011-01-01

    This paper presents a numerical investigation of the thermal and fluid dynamics among the fuel spheres and the cooling fluid, appearing in the core of pebble bed reactor (PBR-Peeble Bed Reactor) using the CFD-Computational Fluid Dynamics CFX 13.0. The paper presents the two analysis results. In the first phase it was considered two heat transfer models for the fuel spheres. In a model it was established volumetric load generation, with thermal conduction for both the fuel and coating. The other model prescribes a heat flux at the sphere surfaces. In this analysis, it was proceed two simulation in the two sphere arrangements, one considering the spheres in contact, and the other with 2 mm spacing between them. At the second analysis it was evaluated the sphere arrangement influence on the thermal and fluid dynamic behavior of the bed. The four simulations present differences in the flow and in the surface and maximum temperature profiles of the coating.(author)

  8. Development of a Reliable Fuel Depletion Methodology for the HTR-10 Spent Fuel Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Kiwhan [Los Alamos National Laboratory; Beddingfield, David H. [Los Alamos National Laboratory; Geist, William H. [Los Alamos National Laboratory; Lee, Sang-Yoon [unaffiliated

    2012-07-03

    A technical working group formed in 2007 between NNSA and CAEA to develop a reliable fuel depletion method for HTR-10 based on MCNPX and to analyze the isotopic inventory and radiation source terms of the HTR-10 spent fuel. Conclusions of this presentation are: (1) Established a fuel depletion methodology and demonstrated its safeguards application; (2) Proliferation resistant at high discharge burnup ({approx}80 GWD/MtHM) - Unfavorable isotopics, high number of pebbles needed, harder to reprocess pebbles; (3) SF should remain under safeguards comparable to that of LWR; and (4) Diversion scenarios not considered, but can be performed.

  9. Source Term Analysis of the Irradiated Graphite in the Core of HTR-10

    Directory of Open Access Journals (Sweden)

    Xuegang Liu

    2017-01-01

    Full Text Available The high temperature gas-cooled reactor (HTGR has potential utilization due to its featured characteristics such as inherent safety and wide diversity of utilization. One distinct difference between HTGR and traditional pressurized water reactor (PWR is the large inventory of graphite in the core acting as reflector, moderator, or structure materials. Some radionuclides will be generated in graphite during the period of irradiation, which play significant roles in reactor safety, environmental release, waste disposal, and so forth. Based on the actual operation of the 10 MW pebble bed high temperature gas-cooled reactor (HTR-10 in Tsinghua University, China, an experimental study on source term analysis of the irradiated graphite has been done. An irradiated graphite sphere was randomly collected from the core of HTR-10 as sample in this study. This paper focuses on the analytical procedure and the establishment of the analytical methodology, including the sample collection, graphite sample preparation, and analytical parameters. The results reveal that the Co-60, Cs-137, Eu-152, and Eu-154 are the major γ contributors, while H-3 and C-14 are the dominating β emitting nuclides in postirradiation graphite material of HTR-10. The distribution profiles of the above four nuclides are also presented.

  10. Graphite for fusion energy applications

    International Nuclear Information System (INIS)

    Eatherly, W.P.; Clausing, R.E.; Strehlow, R.A.; Kennedy, C.R.; Mioduszewski, P.K.

    1987-03-01

    Graphite is in widespread and beneficial use in present fusion energy devices. This report reflects the view of graphite materials scientists on using graphite in fusion devices. Graphite properties are discussed with emphasis on application to fusion reactors. This report is intended to be introductory and descriptive and is not intended to serve as a definitive information source

  11. Study of the thermal drop at the uranium-can interface for fuel elements in gas-graphite reactors; Etude de la chute thermique au contact uranium-gaine pour des elements combustibles de reacteur de la filiere graphite-gaz

    Energy Technology Data Exchange (ETDEWEB)

    Faussat, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Levenes, G; Michel, M [Societe Industrielle de Combustible Nucleaire (France)

    1964-07-01

    The report reviews the tests now under way at the CEA, for determining the thermal contact resistance at the uranium-can interface for fuel elements used in gas-graphite type reactors. These are laboratory tests carried out with equipment based on the principle of a heat flow across a stack of test pieces having planar contact surfaces. The following points emerge from this work: - for a metallic uranium element canned in magnesium, of the type G-2 or EDF-2, a value of 0.2 deg C/W/cm{sup 2} seems reasonable for can temperatures of 400 deg C and above. - this value is independent of the micro-geometric state of the uranium surface in a range of roughness which easily includes those observed on tubes and rods produced industrially. - for the internal cans of elements cooled internally and externally, the value of the contact resistance for temperatures of under 400 deg C as a function of the stresses in the can has not yet been measured exactly. (authors) [French] Le rapport fait le point des essais actuellement en cours au CEA pour determiner la resistance thermique de contact uranium-gaine pour des reacteurs de la filiere graphite-gaz. Ces essais sont effectues en laboratoire sur des appareils bases sur le principe d'une circulation de flux de chaleur a travers un empilement d'eprouvettes dont les faces en contact sont planes. De l'etude, il ressort essentiellement que: - pour un element a uranium metallique et gaine de magnesium type G-2 ou EdF-2, on peut admettre la valeur de 0,2 deg C/W/cm{sup 2} pour des temperatures de gaines de 400 deg C et plus. - cette valeur ne depend pas de l'etat de surface microgeometrique de l'uranium pour un domaine de rugosites couvrant largement celles que l'on observe sur des tubes et barreaux fabriques en serie. - pour les gaines internes d'elements a refroidissement interne et externe la valeur de la resistance de contact reste a preciser pour les temperatures inferieures a 400 deg C, en fonction des contraintes existant dans les

  12. Stable isotope compositions of quartz pebbles and their fluid inclusions as tracers of sediment provenance: Implications for gold- and uranium-bearing quartz pebble conglomerates

    Energy Technology Data Exchange (ETDEWEB)

    Vennemann, T.W.; Kesler, S.E.; O' Neil, J.R. (Univ. of Michigan, Ann Arbor (United States))

    1992-09-01

    Oxygen isotope compositions of pebbles from late Archean to paleo-Proterozoic gold- and/or uranium-bearing oligomictic quartz pebble conglomerates of the Witwatersrand district, South Africa, and Huronian Supergroup, Canada, were determined in an attempt to define the nature of the source terrain. The [delta][sup 18]O values of quartz pebbles within any one sample typically vary by [approximately] 4[per thousand] or more, but occasionally by as much as 8[per thousand], even for adjacent pebbles within the same hand specimen. In addition, adjacent quartz pebbles of widely contrasting [delta][sup 18]O values also preserve distinct isotopic signatures of their fluid inclusions. This overall heterogeneity suggests that the pebbles did not undergo significant oxygen isotope exchange after incorporation in the conglomerates. Therefore, oxygen isotope analyses of such quartz pebbles, in combination with a detailed investigation of their mineral and fluid inclusions, can provide a useful method for characterizing pebble populations and hence dominant sediment source modes. Comparison of values found in this study with [delta][sup 18]O values of quartz from Archean granites, pegmatites, and mesothermal greenstone gold veins, i.e., [delta][sup 18]O values of sources commonly proposed for the conglomerate ores, suggests that uranium is derived from a granitic source, whereas gold has a mesothermal greenstone gold source. Low [delta][sup 18]O values of chert pebbles (9[per thousand] to 11.5[per thousand]) relative to those expected for Archean and Proterozoic marine cherts (commonly [ge] 17[per thousand]) effectively exclude marine cherts, and therefore, auriferous iron formations and exhalatives, as likely sources of gold.

  13. Stable isotope compositions of quartz pebbles and their fluid inclusions as tracers of sediment provenance: Implications for gold- and uranium-bearing quartz pebble conglomerates

    International Nuclear Information System (INIS)

    Vennemann, T.W.; Kesler, S.E.; O'Neil, J.R.

    1992-01-01

    Oxygen isotope compositions of pebbles from late Archean to paleo-Proterozoic gold- and/or uranium-bearing oligomictic quartz pebble conglomerates of the Witwatersrand district, South Africa, and Huronian Supergroup, Canada, were determined in an attempt to define the nature of the source terrain. The δ 18 O values of quartz pebbles within any one sample typically vary by ∼ 4 per-thousand or more, but occasionally by as much as 8 per-thousand, even for adjacent pebbles within the same hand specimen. In addition, adjacent quartz pebbles of widely contrasting δ 18 O values also preserve distinct isotopic signatures of their fluid inclusions. This overall heterogeneity suggests that the pebbles did not undergo significant oxygen isotope exchange after incorporation in the conglomerates. Therefore, oxygen isotope analyses of such quartz pebbles, in combination with a detailed investigation of their mineral and fluid inclusions, can provide a useful method for characterizing pebble populations and hence dominant sediment source modes. Comparison of values found in this study with δ 18 O values of quartz from Archean granites, pegmatites, and mesothermal greenstone gold veins, i.e., δ 18 O values of sources commonly proposed for the conglomerate ores, suggests that uranium is derived from a granitic source, whereas gold has a mesothermal greenstone gold source. Low δ 18 O values of chert pebbles (9 per-thousand to 11.5 per-thousand) relative to those expected for Archean and Proterozoic marine cherts (commonly ≥ 17 per-thousand) effectively exclude marine cherts, and therefore, auriferous iron formations and exhalatives, as likely sources of gold

  14. Initial Comparison of Baseline Physical and Mechanical Properties for the VHTR Candidate Graphite Grades

    Energy Technology Data Exchange (ETDEWEB)

    Carroll, Mark C. [Idaho National Lab. (INL), Idaho Falls, ID (United States). VHTR Program

    2014-09-01

    High-purity graphite is the core structural material of choice in the Very High Temperature Reactor (VHTR) design, a graphite-moderated, helium-cooled configuration capable of producing thermal energy for power generation as well as process heat for industrial applications that require temperatures higher than the outlet temperatures of present nuclear reactors. The Baseline Graphite Characterization Program is establishing accurate as-manufactured mechanical and physical property distributions in nuclear-grade graphites by providing comprehensive data that captures the level of variation in measured values. In addition to providing a thorough comparison between these values in different graphite grades, the program is also carefully tracking individual specimen source, position, and orientation information in order to provide comparisons both in specific properties and in the associated variability between different lots, different billets, and different positions from within a single billet. This report is a preliminary comparison between each of the grades of graphite that are considered “candidate” grades from four major international graphite producers. These particular grades (NBG-18, NBG-17, PCEA, IG-110, and 2114) are the major focus of the evaluations presently underway on irradiated graphite properties through the series of Advanced Graphite Creep (AGC) experiments. NBG-18, a medium-grain pitch coke graphite from SGL from which billets are formed via vibration molding, was the favored structural material in the pebble-bed configuration. NBG-17 graphite from SGL is essentially NBG-18 with the grain size reduced by a factor of two. PCEA, petroleum coke graphite from GrafTech with a similar grain size to NBG-17, is formed via an extrusion process and was initially considered the favored grade for the prismatic layout. IG-110 and 2114, from Toyo Tanso and Mersen (formerly Carbone Lorraine), respectively, are fine-grain grades produced via an isomolding

  15. Reasons of an experimental effort for pebble bed reactors. A program of measurements in the CESAR reactor at Cadarache

    Energy Technology Data Exchange (ETDEWEB)

    Scherer, W; Bock, H J; Krings, F; Neef, R D; Langlet, G; Dixmier, M; Laponche, B; Morier, F

    1972-06-15

    An extended experimental program on neutron physics of HTR fuel balls is being performed in the graphite moderated critical faclity CESAR at CEN Cadarache (France). The experiments are done in the frame of a cooperation between KFA Juelich and CEA Cadarache.

  16. Structural characteristics of a graphite moderated critical assembly for a Zero Power reactor at IEA (Brazil)

    International Nuclear Information System (INIS)

    Almeida Ferreira, A.C. de; Hukai, R.Y.

    1975-01-01

    The structural characteristics of a graphite moderated core of a critical assembly to be installed in the Zero Power Reactor of IEA have been defined. These characteristics are the graphite block dimensions, the number and dimensions of the holes in the graphite, the pitch, the dimensions of the sticks of fuel and graphite to be inserted in the holes, and the mechanical reproducibility of the system. The composition of the fuel and moderator sticks were also defined. The main boundary conditions were the range of the relation C/U and C/TH used in commercial HTGR and the neutronics homogeneity

  17. Recent developments in graphite. [Use in HTGR and aerospace

    Energy Technology Data Exchange (ETDEWEB)

    Cunningham, J.E.

    1983-01-01

    Overall, the HTGR graphite situation is in excellent shape. In both of the critical requirements, fuel blocks and support structures, adequate graphites are at hand and improved grades are sufficiently far along in truncation. In the aerospace field, GraphNOL N3M permits vehicle performance with confidence in trajectories unobtainable with any other existing material. For fusion energy applications, no other graphite can simultaneously withstand both extreme thermal shock and neutron damage. Hence, the material promises to create new markets as well as to offer a better candidate material for existing applications.

  18. Corrosion-induced microstructural changes in a US core graphite

    International Nuclear Information System (INIS)

    Eatherly, W.P.; Lee, D.A.

    1981-01-01

    The results reported here apply to Great Lakes grade H-451 graphite, the core graphite specified for the US HTGR. This graphite is structurally similar to the German reflector grades we have investigated at ORNL, and hence should be applicable to them if similar impurity levels are obtained. Moreover, these results extend and confirm the behavior pattern exhibited by the fuel matrix material A3-3 reported in the previous paper, although the effects are more pronounced in A3-3 presumably due to its resin-type binder and low heat-treatment temperatures

  19. The reprocessing of advanced mixed lithium orthosilicate/metatitanate tritium breeder pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Leys, Oliver, E-mail: oliver.leys@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, Eggenstein-Leopoldshafen, 76344 (Germany); Bergfeldt, Thomas; Kolb, Matthias H.H.; Knitter, Regina [Karlsruhe Institute of Technology, Institute for Applied Materials, Eggenstein-Leopoldshafen, 76344 (Germany); Goraieb, Aniceto A. [Karlsruhe Beryllium Handling Facility, Eggenstein-Leopoldshafen, 76344 (Germany)

    2016-06-15

    Highlights: • The recycling of advanced breeder pebbles without a deterioration of the material properties is possible using a melt-based process. • The only accumulation of impurities upon reprocessing, results from the platinum crucible alloy used for processing. • It is possible to replenish burnt-up lithium by additions of LiOH·H{sub 2}O to the melt during reprocessing. - Abstract: The recycling of tritium breeding materials will be necessary for any future use of nuclear fusion energy due to economical as well as ecological considerations. In the case of the solid breeder blanket concept, the ceramic pebble beds that are intended for the generation of tritium will eventually need to be restored due to depleted lithium levels as well as due to fractured pebbles, which will cause a deterioration of the pebble bed properties. It is proposed that the pebbles, which are fabricated using a melt-based process, are recycled using the same initial process, by replenishing the lithium levels and reforming the pebbles at the same time. To prove this recycling scheme, advanced ceramic pebbles were fabricated and then re-melted multiple times to prove that the reprocessing did not have any negative effect on the pebble properties and secondly, pebbles were produced with a simulated lithium burn-up and subsequently replenished by additions of LiOH to the melt. It was shown that the re-melting and lithium re-enrichment had no effect on the pebble properties, demonstrating that a melt-based process is suitable for recycling used breeder pebbles.

  20. Mechanical behavior of Be–Ti pebbles at blanket relevant temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Kurinskiy, Petr, E-mail: petr.kurinskiy@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials—Applied Materials Physics (IAM-AWP), P.O. Box 3640, 76021 Karlsruhe (Germany); Rolli, Rolf [Karlsruhe Institute of Technology, Institute for Applied Materials—Materials Biomechanics (IAM-WBM), P.O. Box 3640, 76021 Karlsruhe (Germany); Kim, Jae-Hwan; Nakamichi, Masaru [Breeding Functional Materials Development Group, Department of Blanket Fusion Institute, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 2-166 Oaza-Obuchi-Aza-Omotedate, Rokkasho-mura, Kamikita-gun, Aoori 039-3212 (Japan)

    2016-11-01

    Highlights: • Mechanical behavior of two kinds of Be–Ti pebbles in the temperature range of 400–800 °C was investigated. • It was experimentally shown that Be-7 at.%Ti pebbles have the enhanced ductile properties compared to Be-7.7 at.%Ti pebbles. • Brittle failure of both kinds of Be–Ti pebbles was observed by testing at 400 °C using the constant loading with 150 N. - Abstract: Mechanical performance of beryllium-based materials is a matter of a great interest from the point of view of their use as neutron multipliers of the tritium breeding blankets. The compression strains which can occur in beryllium pebble beds under blanket working conditions will lead to deformation or even failure of individual pebbles [1,2] (Reimann et al. 2002; Ishitsuka and Kawamura, 1995). Mechanical behavior of Be–Ti pebbles having chemical contents of Be-7.0 at.% Ti and Be-7.7 at.%Ti was investigated in the temperature range of 400–800 °C. Constant loads varying from 10 up to 150 N were applied uniaxially. It was shown that Be–Ti pebbles compared to pure beryllium pebbles possess much lower ductility, although their strength properties exceed corresponding characteristics of pure beryllium. Also, the influence of titanium content on mechanical behavior of Be–Ti pebbles was investigated. Specific features of deformation of pure beryllium and Be–Ti pebbles having different titanium contents at blanket operation temperatures are discussed.

  1. Fabrication of Li_2TiO_3 pebbles by a selective laser sintering process

    International Nuclear Information System (INIS)

    Zhou, Qilai; Gao, Yue; Liu, Kai; Xue, Lihong; Yan, Youwei

    2015-01-01

    Highlights: • Selective laser sintering (SLS) is employed to fabricate ceramic pebbles. • Quantities and diameter of the pebbles could be easily controlled by adjusting the model of pebbles. • All the pebbles could be prepared at a time within several minutes. • The Li_2TiO_3 pebbles sintered at 1100 °C show a notable crush load of 43 N. - Abstract: Lithium titanate, Li_2TiO_3, is an important tritium breeding material for deuterium (D)–tritium (T) fusion reactor. In test blanket module (TBM) design of China, Li_2TiO_3 is considered as one candidate material of tritium breeders. In this study, selective laser sintering (SLS) technology was introduced to fabricate Li_2TiO_3 ceramic pebbles. This fabrication process is computer assisted and has a high level of flexibility. Li_2TiO_3 powder with a particle size of 1–3 μm was used as the raw material, whilst epoxy resin E06 was adopted as a binder. Green Li_2TiO_3 pebbles with certain strengths were successfully prepared via SLS. Density of the green pebbles was subsequently increased by cold isostatic pressing (CIP) process. Li_2TiO_3 pebbles with a diameter of about 2 mm were obtained after high temperature sintering. Density of the pebbles reaches 80% of theoretical density (TD) with a comparable crush load of 43 N. This computer assisted approach provides a new efficient route for the production of Li_2TiO_3 ceramic pebbles.

  2. Carbon-14 Graphitization Chemistry

    Science.gov (United States)

    Miller, James; Collon, Philippe; Laverne, Jay

    2014-09-01

    Accelerator Mass Spectrometry (AMS) is a process that allows for the analysis of mass of certain materials. It is a powerful process because it results in the ability to separate rare isotopes with very low abundances from a large background, which was previously impossible. Another advantage of AMS is that it only requires very small amounts of material for measurements. An important application of this process is radiocarbon dating because the rare 14C isotopes can be separated from the stable 14N background that is 10 to 13 orders of magnitude larger, and only small amounts of the old and fragile organic samples are necessary for measurement. Our group focuses on this radiocarbon dating through AMS. When performing AMS, the sample needs to be loaded into a cathode at the back of an ion source in order to produce a beam from the material to be analyzed. For carbon samples, the material must first be converted into graphite in order to be loaded into the cathode. My role in the group is to convert the organic substances into graphite. In order to graphitize the samples, a sample is first combusted to form carbon dioxide gas and then purified and reduced into the graphite form. After a couple weeks of research and with the help of various Physics professors, I developed a plan and began to construct the setup necessary to perform the graphitization. Once the apparatus is fully completed, the carbon samples will be graphitized and loaded into the AMS machine for analysis.

  3. Melting temperature of graphite

    International Nuclear Information System (INIS)

    Korobenko, V.N.; Savvatimskiy, A.I.

    2001-01-01

    Full Text: Pulse of electrical current is used for fast heating (∼ 1 μs) of metal and graphite specimens placed in dielectric solid media. Specimen consists of two strips (90 μm in thick) placed together with small gap so they form a black body model. Quasy-monocrystal graphite specimens were used for uniform heating of graphite. Temperature measurements were fulfilled with fast pyrometer and with composite 2-strip black body model up to melting temperature. There were fulfilled experiments with zirconium and tungsten of the same black body construction. Additional temperature measurements of liquid zirconium and liquid tungsten are made. Specific heat capacity (c P ) of liquid zirconium and of liquid tungsten has a common feature in c P diminishing just after melting. It reveals c P diminishing after melting in both cases over the narrow temperature range up to usual values known from steady state measurements. Over the next wide temperature range heat capacity for W (up to 5000 K) and Zr (up to 4100 K) show different dependencies of heat capacity on temperature in liquid state. The experiments confirmed a high quality of 2-strip black body model used for graphite temperature measurements. Melting temperature plateau of tungsten (3690 K) was used for pyrometer calibration area for graphite temperature measurement. As a result, a preliminary value of graphite melting temperature of 4800 K was obtained. (author)

  4. Glassy carbon coated graphite for nuclear applications

    International Nuclear Information System (INIS)

    Delpeux S; Cacciaguerra T; Duclaux L

    2005-01-01

    Taking into account the problems caused by the treatment of nuclear wastes, the molten salts breeder reactors are expected to a great development. They use a molten fluorinated salt (mixture of LiF, BeF 2 , ThF 4 , and UF 4 ) as fuel and coolant. The reactor core, made of graphite, is used as a neutrons moderator. Despite of its compatibility with nuclear environment, it appears crucial to improve the stability and inertness of graphite against the diffusion of chemicals species leading to its corrosion. One way is to cover the graphite surface by a protective impermeable deposit made of glassy carbon obtained by the pyrolysis of phenolic resin or polyvinyl chloride precursors. The main difficulty in the synthesis of glassy carbon is to create exclusively, in the primary pyrolysis product, a micro-porosity of about twenty Angstroms which closes later at higher temperature. Therefore, the evacuation of the volatile products occurring mainly between 330 and 600 C, must progress slowly to avoid the material to crack. In this study, the optimal parameters for the synthesis of glassy carbon as well as glassy carbon deposits on nuclear-type graphite pieces are discussed. Both thermal treatment of phenolic and PVC resins have been performed. The structure and micro-texture of glassy carbon have been investigated by X-ray diffraction, scanning and transmission electron microscopies and helium pycno-metry. Glassy carbon samples (obtained at 1200 C) show densities ranging from 1.3 to 1.55 g/cm 3 and closed pores with nano-metric size (∼ 5 to 10 nm) appear clearly on the TEM micrographs. Then, a thermal treatment to 2700 C leads to the shrinkage of the entangled graphene ribbons, in good agreement with the proposed texture model for glassy carbon. Glassy carbon deposits on nuclear graphite have been developed by an impregnation method. The uniformity of the deposit depends clearly on the surface texture and the chemistry of the graphite substrate. The deposit regions where

  5. Glassy carbon coated graphite for nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Delpeux, S.; Cacciaguerra, T.; Duclaux, L. [Orleans Univ., CRMD, CNRS, 45 (France)

    2005-07-01

    Taking into account the problems caused by the treatment of nuclear wastes, the molten salts breeder reactors are expected to a great development. They use a molten fluorinated salt (mixture of LiF, BeF{sub 2}, ThF{sub 4}, and UF{sub 4}) as fuel and coolant. The reactor core, made of graphite, is used as a neutrons moderator. Despite of its compatibility with nuclear environment, it appears crucial to improve the stability and inertness of graphite against the diffusion of chemicals species leading to its corrosion. One way is to cover the graphite surface by a protective impermeable deposit made of glassy carbon obtained by the pyrolysis of phenolic resin [1,2] or polyvinyl chloride [3] precursors. The main difficulty in the synthesis of glassy carbon is to create exclusively, in the primary pyrolysis product, a micro-porosity of about twenty Angstroms which closes later at higher temperature. Therefore, the evacuation of the volatile products occurring mainly between 330 and 600 C, must progress slowly to avoid the material to crack. In this study, the optimal parameters for the synthesis of glassy carbon as well as glassy carbon deposits on nuclear-type graphite pieces are discussed. Both thermal treatment of phenolic and PVC resins have been performed. The structure and micro-texture of glassy carbon have been investigated by X-ray diffraction, scanning and transmission electron microscopies and helium pycno-metry. Glassy carbon samples (obtained at 1200 C) show densities ranging from 1.3 to 1.55 g/cm{sup 3} and closed pores with nano-metric size ({approx} 5 to 10 nm) appear clearly on the TEM micrographs. Then, a thermal treatment to 2700 C leads to the shrinkage of the entangled graphene ribbons (Fig 1), in good agreement with the proposed texture model for glassy carbon (Fig 2) [4]. Glassy carbon deposits on nuclear graphite have been developed by an impregnation method. The uniformity of the deposit depends clearly on the surface texture and the chemistry

  6. Physics experiments in graphite lattices (1962); Experiences sur les reseaux a graphite (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Bacher, P; Cogne, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1962-07-01

    A review is made of the various experimental methods used to determine the physics of graphite, natural uranium lattices: integral lattice experiments; both absolute and differential, effective cross section measurements, both by activation methods and by analysis of irradiated fuels, fine structure measurements. A number of experimental results are also given. (authors) [French] On decrit les differentes methodes experimentales utilisees pour determiner les parametres physiques de reseaux a uranium-graphite. Il s'agit d'experiences globales: mesures absolues et relatives de laplaciens, mesures de sections efficaces effectives par activation et par analyses de combustibles irradies, mesures de structures fines. Un certain nombre de resultats experimentaux sont communiques. (auteurs)

  7. Graphite-moderated and heavy water-moderated spectral shift controlled reactors

    International Nuclear Information System (INIS)

    Alcala Ruiz, F.

    1984-01-01

    It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs

  8. Planet population synthesis driven by pebble accretion in cluster environments

    Science.gov (United States)

    Ndugu, N.; Bitsch, B.; Jurua, E.

    2018-02-01

    The evolution of protoplanetary discs embedded in stellar clusters depends on the age and the stellar density in which they are embedded. Stellar clusters of young age and high stellar surface density destroy protoplanetary discs by external photoevaporation and stellar encounters. Here, we consider the effect of background heating from newly formed stellar clusters on the structure of protoplanetary discs and how it affects the formation of planets in these discs. Our planet formation model is built on the core accretion scenario, where we take the reduction of the core growth time-scale due to p