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Sample records for breeder pebble beds

  1. Numerical simulation of ceramic breeder pebble bed thermal creep behavior

    International Nuclear Information System (INIS)

    The evolution of ceramic breeder pebble bed thermal creep deformation subjected to an external load and a differential thermal stress was studied using a modified discrete numerical code previously developed for the pebble bed thermomechanical evaluation. The rate change of creep deformation was modeled at the particle contact based on a diffusion creep mechanism. Numerical results of strain histories have compared reasonably well with those of experimentally observed data at 740 C using activation energy of 180 KJ/mole. Calculations also show that, at this activation energy level, a particle bed at an elevated temperature of 800 C may cause undesired local sintering at a later time when it is subjected to an external load of 6.3 MPa. Thus, by tracking the stress histories inside a breeder pebble bed the numerical simulation provides an indication of whether the bed may encounter an undesired condition under a typical operating condition. (orig.)

  2. Numerical simulation of ceramic breeder pebble bed thermal creep behavior

    International Nuclear Information System (INIS)

    The evolution of ceramic breeder pebble bed thermal creep deformation subjected to an external load and a differential thermal stress was studied using a modified discrete numerical code previously developed for the pebble bed thermomechanical evaluation. The rate change of creep deformation was modeled at the particle contact based on a diffusion creep mechanism. Numerical results of strain histories have shown lower values as compared to those of experimentally observed data at 740 deg. C using an activation energy of 180 kJ/mol. Calculations also show that, at this activation energy level, a particle bed at an elevated temperature of 800 deg. C may cause too much particle overlapping with a contact radius growth beyond 0.65 radius at a later time, when it is subjected to an external load of 6.3 MPa. Thus, by tracking the stress histories inside a breeder pebble bed the numerical simulation provides an indication of whether the bed may encounter an undesired condition under a typical operating condition

  3. Particle flow of ceramic breeder pebble beds in bi-axial compression experiments

    International Nuclear Information System (INIS)

    Pebble beds of Tritium breeding ceramic material are investigated within the framework of developing solid breeder blankets for future nuclear fusion power plants. For the thermo-mechanical characterisation of such pebble beds, bed compression experiments are the standard tools. New bi-axial compression experiments on 20 and 30 mm high pebble beds show pebble flow effects much more pronounced than in previous 10 mm beds. Owing to the greater bed height, conditions are reached where the bed fails in cross direction and unhindered flow of the pebbles occurs. The paper presents measurements for the orthosilicate and metatitanate breeder materials that are envisaged to be used in a solid breeder blanket. The data are compared with calculations made with a Drucker-Prager soil model within the finite-element code ABAQUS, calibrated with data from other experiments. It is investigated empirically whether internal bed friction angles can be determined from pebble beds of the considered heights, which would simplify, and broaden the data base for, the calibration of the Drucker-Prager pebble bed models

  4. Plutonium destruction with pebble bed type HTGRs using Pu burner balls and breeder balls

    International Nuclear Information System (INIS)

    It was made clear that pebble bed type HTGRs using Pu burner balls (pu balls) and breeder balls (Th balls) possesses a potential to burn weapons-grade Pu to 740 Gwd/TPu. The total amounts of Pu and 239Pu of can reduced to about 20 and 1%, respectively. (author). 10 refs, 4 figs, 2 tabs

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

    International Nuclear Information System (INIS)

    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

  6. Disposition of weapon-grade plutonium with pebble bed type HTGRs using Pu burner balls and Th breeder balls

    International Nuclear Information System (INIS)

    A concept of reactor system was developed with which weapons-grade plutonium could be made perfectly worthless in use for weapons. It is a pebble bed type HTGR using Pu burner ball fuels and Th breeder ball fuels. The residual amounts of 239Pu in spent Pu balls become less than 1% of the initial loading. Furthermore, a method was found that the power coefficient could be made negative by heavy Pu loading in the Pu burner ball fuels

  7. Reactivity control system of a passively safe thorium breeder pebble bed reactor

    International Nuclear Information System (INIS)

    Highlights: • A worth of over 15,000 pcm ensures achieving long-term cold shutdown in thorium PBR. • Control rod worth in side reflector is insufficient due to low-power breeder zone. • 20 control rods, just outside the driver zone, can achieve long-term cold shutdown. • BF3 gas can be inserted for reactor shutdown, but only in case of emergency. • Perturbation theory accurately predicts absorber gas worth for many concentrations. - Abstract: This work investigates the neutronic design of the reactivity control system for a 100 MWth passively safe thorium breeder pebble bed reactor (PBR), a conceptual design introduced previously by the authors. The thorium PBR consists of a central driver zone of 100 cm radius, surrounded by a breeder zone with 300 cm outer radius. The fissile content of the breeder zone is low, leading to low fluxes in the radial reflector region. Therefore, a significant decrease of the control rod worth at this position is anticipated. The reactivity worth of control rods in the side reflector and at alternative in-core positions is calculated using different techniques, being 2D neutron diffusion, perturbation theory and more accurate 3D Monte Carlo models. Sensitivity coefficients from perturbation theory provide a first indication of effective control rod positions, while the 2D diffusion models provide an upper limit on the reactivity worth achievable at a certain radial position due to the homogeneous spreading of the absorber material over the azimuthal domain. Three dimensional forward calculations, e.g. in KENO, are needed for an accurate calculation of the total control rod worth. The two dimensional homogeneous calculations indicate that the reactivity worth in the radial reflector is by far insufficient to achieve cold reactor shutdown, which requires a control rod worth of over 15 000 pcm. Three dimensional heterogeneous KENO calculations show that placing 20 control rods just outside the driver channel, between 100 cm and

  8. Analysis of the running-in phase of a Passively Safe Thorium Breeder Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Highlights: • This work analyzes important trends of the running-in phase of a thorium breeder PBR. • Depletion equations are solved for important actinides and a fission product pair. • Breeding U-233 is achieved in 7 years by cleverly adjusting the feed fuel enrichment. • A safety analysis shows the thorium PBR is passively safe during the running-in phase. - Abstract: The present work investigates the running-in phase of a 100 MWth Passively Safe Thorium Breeder Pebble Bed Reactor (PBR), a conceptual design introduced in previous equilibrium core design studies by the authors. Since U-233 is not available in nature, an alternative fuel, e.g. U-235/U-238, is required to start such a reactor. This work investigates how long it takes to converge to the equilibrium core composition and to achieve a net production of U-233, and how this can be accelerated. For this purpose, a fast and flexible calculation scheme was developed to analyze these aspects of the running-in phase. Depletion equations with an axial fuel movement term are solved in MATLAB for the most relevant actinides (Th-232, Pa-233, U-233, U-234, U-235, U-236 and U-238) and the fission products are lumped into a fission product pair. A finite difference discretization is used for the axial coordinate in combination with an implicit Euler time discretization scheme. Results show that a time dependent adjustment scheme for the enrichment (in case of U-235/U-238 start-up fuel) or U-233 weight fraction of the feed driver fuel helps to restrict excess reactivity, to improve the fuel economy and to achieve a net production of U-233 faster. After using U-235/U-238 startup fuel for 1300 days, the system starts to work as a breeder, i.e. the U-233 (and Pa-233) extraction rate exceeds the U-233 feed rate, within 7 years after start of reactor operation. The final part of the work presents a basic safety analysis, which shows that the thorium PBR fulfills the same passive safety requirements as the

  9. Destruction of weapons-grade plutonium with pebble bed type HTGRs using burner balls and breeder balls

    International Nuclear Information System (INIS)

    As the method of disposing the plutonium coming from disassembled weapons, the method of burning the fuel in which the plutonium is mixed with a parent material in LWRs or the disposal by glass solidification is proposed. In the former method, it is desirable to do the reprocessing of spent fuel for effectively utilizing fission products. The latter method needs watch against the diversion of the plutonium. The authors devised the method of effectively annihilating plutonium by separating into the burner balls of plutonium and the breeder balls of a parent material, and burning those by mixing in a pebble bed type high temperature gas-cooled reactor, while continuously exchanging them. It was clarified from the aspect of nuclear characteristics that by using this method, 239Pu can be annihilated to the state of enabling the direct abandonment without reprocessing. The flow of burner balls and breeder balls in the reactor is shown, and multi-pass fuel exchange method was adopted to burn Pu in burner balls up. The rate of Pu annihilation was determined by the change of the amount of Pu for the burnup evaluated by lattice burning calculation. The maximum amount of Pu charge in one burner ball is limited by the maximum allowable power output of burner balls. (K.I.)

  10. Pebble bed pebble motion: Simulation and Applications

    CERN Document Server

    Cogliati, Joshua J

    2011-01-01

    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, P...

  11. Pebble bed packing in prismatic containers

    International Nuclear Information System (INIS)

    Highlights: • The essential part of ceramic breeder blankets is pebble beds. • The packing factor for blanket relevant cavities must be known. • Tomography experiments revealed details of packing arrangements. • Packing experiments confirm that reference packing factors will be achieved. -- Abstract: New analyses of previous tomography investigations show in detail void fraction fluctuations close to walls generated by regular pebble arrangements. Local packing factors within the pebble bed were determined for characteristic zones. These results are very helpful for the interpretation of the packing experiments performed with spherical pebbles in different kinds of Plexiglas containers dominated by flat walls. The packing factors for single-size pebbles in the containers with a piston on top are fairly independent of bed height unless the height to diameter ratio becomes less than 10. For the closed rectangular containers, the development of structured packings is rendered more difficult. However, for blanket relevant bed heights, both for orthosilicate and beryllium pebbles, the packing factors obtained which agree well with previously obtained reference values

  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. Reactor vessel for pebble beds

    International Nuclear Information System (INIS)

    The wall and the bottom of the vessel for the gas-cooled pebble-bed reactor consist of numerous blocks of graphite or carbon rock piled up. They are held together by an exterior cylindrical or polygonal ring and supported by a foundation. The blocks form coherent sectors resp. annular sectors with well-defined separating lines. At high temperatures or load change operation these sectors behave like monolithic blocks expanding heely and contracting again, the center of the vessel remaining fixed. The forces causing the compression result from the own weight of the sectors and the weight of the pebble bed. This motion is supported by the convex arrangement of the opposite surfaces of the sectors and the supporting walls and by roller bearings. The bottom of the vessel may be designed funnel-shaped, in this way facilitating the removal of spheres. (DG)

  14. Pebble bed modular reactor (PBMR)

    International Nuclear Information System (INIS)

    In 1993, the pebble bed modular reactor (PBMR) was identified by ESKOM, the electric utility of South Africa, as a leading option for the installation of new generating capacity to their electric grid. This innovative nuclear power plant incorporates a closed cycle primary coolant system utilizing helium to transport heat energy directly from the modular pebble bed reactor to a recuperative power conversion unit with a single-shaft turbine/compressor/generator. This replacement of the steam cycle that is common in present nuclear power plants (NPP) with a direct gas cycle provides the benefits of simplification and a substantial increase in overall system efficiency with the attendant lowering of capital and operational costs. Although the historical development of this plant is interrelated to other types of high temperature gas cooled reactors (HTGRs), the principle focus herein is on the pebble bed (spherical) fuel element type reactor. The long-term development of this reactor type began in Germany by the KFA Nuclear Research Center (now FZJ). Two pebble bed plants were constructed in Germany, the 46 MW(th)/15 MW(e) Arbeitsgemeinshaft Versuchsreaktor (AVR) and the 750 MW(th)/296 MW(e) thorium high temperature reactor (THTR-300). Basically, these steam/electric plants validated the temperature and fission product retention capabilities of the ceramic (TRISO) coated fuel particle and the safety characteristics of the HTGR. Most notable of the operational achievements was with the AVR in sustaining longterm operation at an average core outlet temperature of 950 deg. C, and in demonstration of safety such as extended loss of forced cooling on the core. More details on the AVR and THTR-300 plants are provided The next evolution of the pebble bed plant began in the early 1980s with development of the modular reactor. This small reactor added the unique characteristic of being able to cool the core entirely by passive heat transfer mechanisms following postulated

  15. Pebble Bed Reactor Dust Production Model

    International Nuclear Information System (INIS)

    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

  16. Progress of R and D on the technology of In-pile irradiation and tritium In-situ extraction experiment of solid breeder pebble bed for CN HCCB in CARR

    International Nuclear Information System (INIS)

    The progress of the key technology of the In-Pile Irradiation and Tritium In-Situ Extraction (IPITISE) experiment was introduced. According to the design and requirements of the Helium cooled ceramic breeder (HCCB) tritium system, the scheme of the IPITISE experiment was established. The primary components of this apparatus included pebble bed assembly (PBA), tritium extraction system (TES), and tritium measurement system (TMS). The primary design and calculation of the structure, nuclear physics, and thermo-hydraulics of the PBA were carried out. It can be concluded that the max weight load of the PBA was 150 g above. The effective thermal neutron flux rate of the PBA would be high as 1.73 E +14 n/cm2 · s. The power of irradiation heat could be reached 18.47 W/cm3. After optimization of the design and parameter of PBA structure, the temperature of reduced activation ferrite/martensite (RAFM) steel with tritium permeation barrier (TPB) coating and Li4SiO4 could be respective controlled under 150∼550℃ and 200∼ 920℃. The in-situ tritium release behaviors would be studied by this experiment as well as the tritium permeation through the structure materials under irradiation condition or the reactor was shut down. Consequently, the irradiation performance of the key materials, the retention characteristics and release behaviors of Li4SiO4 ceramic breeder pebbles, the tritium permeation data of RAFM with TPB coating and credible evaluation of in-situ tritium extraction technology would be provided for China tritium breeder test blanket module in the CIPITISE experiment. (authors)

  17. Reprocessing of lithium titanate pebbles by graphite bed method

    International Nuclear Information System (INIS)

    Lithium titanate enriched by 6Li isotope is considered as a candidate of tritium breeding materials for fusion reactors due to its excellent performance. The reuse of burned Li2TiO3 pebbles is an important issue because of the high costs of 6Li-enriched materials and waste considerations. For this purpose, reprocessing of Li2TiO3 pebbles by graphite bed method was developed. Simulative Li2TiO3 pebbles with low-lithium content according to the expected lithium burn-up were fabricated. After that, Li2TiO3 pebbles were re-fabricated with lithium carbonate as lithium additives, in order to gain the composition of lithium titanate with a Li/Ti ratio of 2. The process was optimized to obtain reprocessed Li2TiO3 pebbles that were suitable for reuse as ceramic breeder. Density, porosity, grain size and crushing load of the reprocessed pebbles were characterized. This process did not deteriorate the properties of the reprocessed pebbles and was almost no waste generation

  18. Absorber rod for pebble-bed reactor

    International Nuclear Information System (INIS)

    The absorber rod that can be moved into the pebble bed from the top reflector is enclosed by a cladding tube which, if it is completely moved down, ends above the pebble bed and is open at the bottom. Through the cladding tube the absorber rod is cooled with gas. The cladding tube consists of e.g. boron steel. If the absorber rod is drawn it takes along the cladding tube which is moved into the guide tube like a telescope. The rigidity of that part of the absorber rod projecting from the pebble bed is thus guaranteed. (DG)

  19. Pressurizing Behavior on Ingress of Coolant into Pebble Bed of Blanket of Fusion DEMO Reactor

    International Nuclear Information System (INIS)

    Solid breeder blankets are being developed as candidate blankets for the Fusion DEMO reactor in Japan. JAEA is performing the development of the water cooled and helium cooled solid breeder blankets. The blanket utilizes ceramic breeder pebbles and multiplier pebbles beds cooled by high pressure water or high pressure helium in the cooling tubes placed in the blanket box structure. In the development of the blanket, it is very important to incorporate the safety technology as well as the performance improvement on tritium production and energy conversion. In the safety design and technology, coolant ingress in the blanket box structure is one of the most important events as the initiators. Especially the thermal hydraulics in the pebble bed in the case of the high pressure coolant ingress is very important to evaluate the pressure propagation and coolant flow behavior. This paper presents the preliminary results of the pressure loss characteristics by the coolant ingress in the pebble bed. Experiments have been performed by using alumina pebble bed (4 litter maximum volume of the pebble bed) and nitrogen gas to simulate the helium coolant ingress into breeder and multiplier pebble beds. Reservoir tank of 10 liter is filled with 1.0 MPa nitrogen. The nitrogen gas is released at the bottom part of the alumina pebble bed whose upper part is open to the atmosphere. The pressure change in the pebble bed is measured to identify the pressure loss. The measured values are compared with the predicted values by Ergun's equation, which is the correlation equation on pressure loss of the flow through porous medium. By the results of the experiments with no constraint on the alumina pebble bed, it was clarified that the measured value agreed in the lower flow rate. However, in the higher flow rate where the pressure loss is high, the measured value is about half of the predicted value. The differences between the measured values and the predicted values will be discussed from

  20. Modelling of thermal and mechanical behaviour of pebble beds

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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. Use of plutonium in pebble bed HTGRs

    International Nuclear Information System (INIS)

    This paper provides a summary of the current status of world-wide inventories of weapon-grade plutonium and plutonium from reprocessing of power reactor fuel. It addresses the use of pebble bed HTGRs for consumption of the plutonium in terms of the fuel cycle options. The requirements and neutronics aspects, and results from parameter studies conducted using pebble bed reactor types, are discussed, along with proliferation and waste disposal aspects. (author)

  3. Numerical simulation of nuclear pebble bed configurations

    Energy Technology Data Exchange (ETDEWEB)

    Shams, A., E-mail: shams@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Roelofs, F., E-mail: roelofs@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Komen, E.M.J., E-mail: komen@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Baglietto, E., E-mail: emiliob@MIT.EDU [Massachusetts Institute of Technology (MIT) (United States)

    2015-08-15

    Highlights: • Numerical simulations of a single face cubic centred pebble bed are performed. • Wide range of turbulence modelling techniques are used to perform these calculations. • The methods include 1-DNS, 1-LES, 3-Hybrid (RANS/LES) and 3-RANS models, respectively. • The obtained results are extensively compared to provide guidelines for such flow regimes. • These guidelines are used to perform reference LES for a limited sized random pebble bed. - Abstract: High Temperature Reactors (HTRs) are being considered all over the world. An HTR uses helium gas as a coolant, while the moderator function is taken up by graphite. The fuel is embedded in the graphite moderator. A particular inherent safety advantage of HTR designs is that the graphite can withstand very high temperatures, that the fuel inside will stay inside the graphite pebble and cannot escape to the surroundings even in the event of loss of cooling. Generally, the core can be designed using a graphite pebble bed. Some experimental and demonstration reactors have been operated using a pebble bed design. The test reactors have shown safe and efficient operation, however questions have been raised about possible occurrence of local hot spots in the pebble bed which may affect the pebble integrity. Analysis of the fuel integrity requires detailed evaluation of local heat transport phenomena in a pebble bed, and since such phenomena cannot easily be modelled experimentally, numerical simulations are a useful tool. As a part of a European project, named Thermal Hydraulics of Innovative Nuclear Systems (THINS), a benchmarking quasi-direct numerical simulation (q-DNS) of a well-defined pebble bed configuration has been performed. This q-DNS will serve as a reference database in order to evaluate the prediction capabilities of different turbulence modelling approaches. A wide range of numerical simulations based on different available turbulence modelling approaches are performed and compared with

  4. Numerical simulation of nuclear pebble bed configurations

    International Nuclear Information System (INIS)

    Highlights: • Numerical simulations of a single face cubic centred pebble bed are performed. • Wide range of turbulence modelling techniques are used to perform these calculations. • The methods include 1-DNS, 1-LES, 3-Hybrid (RANS/LES) and 3-RANS models, respectively. • The obtained results are extensively compared to provide guidelines for such flow regimes. • These guidelines are used to perform reference LES for a limited sized random pebble bed. - Abstract: High Temperature Reactors (HTRs) are being considered all over the world. An HTR uses helium gas as a coolant, while the moderator function is taken up by graphite. The fuel is embedded in the graphite moderator. A particular inherent safety advantage of HTR designs is that the graphite can withstand very high temperatures, that the fuel inside will stay inside the graphite pebble and cannot escape to the surroundings even in the event of loss of cooling. Generally, the core can be designed using a graphite pebble bed. Some experimental and demonstration reactors have been operated using a pebble bed design. The test reactors have shown safe and efficient operation, however questions have been raised about possible occurrence of local hot spots in the pebble bed which may affect the pebble integrity. Analysis of the fuel integrity requires detailed evaluation of local heat transport phenomena in a pebble bed, and since such phenomena cannot easily be modelled experimentally, numerical simulations are a useful tool. As a part of a European project, named Thermal Hydraulics of Innovative Nuclear Systems (THINS), a benchmarking quasi-direct numerical simulation (q-DNS) of a well-defined pebble bed configuration has been performed. This q-DNS will serve as a reference database in order to evaluate the prediction capabilities of different turbulence modelling approaches. A wide range of numerical simulations based on different available turbulence modelling approaches are performed and compared with

  5. Mechanics of the pebble bed reactor

    International Nuclear Information System (INIS)

    In a survey, the quite different type of problems which arise for the reactor designer from the mechanics of the pebble bed are demonstrated by examples. It becomes clear why the apparently simple system of a static heap of pebbles of the same diameter is such a complex problem, so that even after research and development work extending over three decades, it cannot be regarded as completely solved. (orig.)

  6. Preliminary safety analysis of a thorium high-conversion pebble bed reactor

    International Nuclear Information System (INIS)

    An inherently safe thorium High-Conversion Pebble Bed Reactor would combine the inherent safety characteristics of the Pebble Bed Reactor with the favourable waste characteristics and resource availability of the thorium fuel cycle. Previous work by the authors showed that high conversion ratio's can be achieved within a thorium Pebble Bed Reactor (PBR) at a practical operating regime. The thorium PBR core design consists of a cylindrical core with a central driver zone surrounded by a breeder zone. The breeder pebbles have a 30 g heavy metal (HM) loading to enhance conversion of Th-232 into U-233, while the driver pebbles (10 w% U-233) contain a lower metal loading to enhance fission. In previous studies, thorium PBR designs were presented for three core diameters, using a 7.5 g heavy metal (HM) loading for the driver pebbles. The current paper investigates the safety of these thorium PBR designs in terms of reactivity coefficients and possible reactivity insertion due to water ingress. Early results indicated that the values of the reactivity coefficients for the three designs with 7.5 g HM loading per driver pebble were rather small and the possible reactivity insertion due to water ingress was very large. Therefore, also a lower HM loading per driver pebble (4 g) was investigated to reduce the impact of water ingress, since the core becomes less under-moderated. For the three core diameters investigated, it is shown that reducing the metal loading in the driver pebbles to 4 g is indeed advantageous in terms of safety, water ingress leads to a smaller reactivity increase but also the reactivity coefficients become stronger negative. Secondly, the breeding performance of the cores with a 4 g driver pebble HM loading improves. On the downside, the driver pebble residence times become shorter, which could increase fuel reprocessing costs. Fuel pebbles would have to be recycled at an increased rate, which might be more challenging from a practical perspective

  7. Multiscale Analysis of Pebble Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hans Gougar; Woo Yoon; Abderrafi Ougouag

    2010-10-01

    – The PEBBED code was developed at the Idaho National Laboratory for design and analysis of pebble-bed high temperature reactors. The diffusion-depletion-pebble-mixing algorithm of the original PEBBED code was enhanced through coupling with the THERMIX-KONVEK code for thermal fluid analysis and by the COMBINE code for online cross section generation. The COMBINE code solves the B-1 or B-3 approximations to the transport equation for neutron slowing down and resonance interactions in a homogeneous medium with simple corrections for shadowing and thermal self-shielding. The number densities of materials within specified regions of the core are averaged and transferred to COMBINE from PEBBED for updating during the burnup iteration. The simple treatment of self-shielding in previous versions of COMBINE led to inaccurate results for cross sections and unsatisfactory core performance calculations. A new version of COMBINE has been developed that treats all levels of heterogeneity using the 1D transport code ANISN. In a 3-stage calculation, slowing down is performed in 167 groups for each homogeneous subregion (kernel, particle layers, graphite shell, control rod absorber annulus, etc.) Particles in a local average pebble are homogenized using ANISN then passed to the next (pebble) stage. A 1D transport solution is again performed over the pebble geometry and the homogenized pebble cross sections are passed to a 1-d radial model of a wedge of the pebble bed core. This wedge may also include homogeneous reflector regions and a control rod region composed of annuli of different absorbing regions. Radial leakage effects are therefore captured with discrete ordinates transport while axial and azimuthal effects are captured with a transverse buckling term. In this paper, results of various PBR models will be compared with comparable models from literature. Performance of the code will be assessed.

  8. Thorium utilization in a pebble bed reactor

    International Nuclear Information System (INIS)

    Thorium reserves in the earth's crust are much more than those of uranium, which today measure about 1.5 million tonnes of reasonably assured resources, plus 3 million tonnes of estimated additional resources. These large amount of thorium reserves, also available in Turkey encourages to focus on the utilization of thorium. The most remarkable applications of the use of thorium have been in high temperature reactors. The high temperature pebble bed reactor, which has been chosen as the basis for this study, is a close approximation of the thorium utilizing German reactor THTR. Pebble bed reactors have some unique features which are suitable to burn thorium. (i) The fuel is loaded in the form of coated particles, which are embedded in the graphite matrix of the fuel pebbles, allowing exceptionally high heavy metal burnups; and (ii) the continuous (on-line) fuel loading allows a high utilization factor. The criticality search of the pebble bed reactor is computed by the use of the SCALE4.4 code, CSASIX and KENOVa modules. And the in-core fuel management is computed via SCALE4.4 code, ORIGEN-S module

  9. Characterization of the thermal conductivity for ceramic pebble beds

    Science.gov (United States)

    Lo Frano, R.; Aquaro, D.; Scaletti, L.; Olivi, N.

    2015-11-01

    The evaluation of the thermal conductivity of breeder materials is one of the main goals to find the best candidate material for the fusion reactor technology. The aim of this paper is to evaluate experimentally the thermal conductivity of a ceramic material by applying the hot wire method at different temperatures, ranging from 50 to about 800°C. The updated experimental facility, available at the Department of Civil and Industrial Engineering (DICI) of the University of Pisa, used to determine the thermal conductivity of a ceramic material (alumina), will be described along with the measurement acquisition system. Moreover it will be also provided an overview of the current state of art of the ceramic pebble bed breeder thermos-mechanics R&D (e.g. Lithium Orthosilicate (Li4SiO4) and Lithium Metatitanate (Li2TiO3)) focusing on the up-to-date analysis. The methodological approach adopted is articulated in two phase: the first one aimed at the experimental evaluation of thermal conductivity of a ceramic material by means of hot wire method, to be subsequently used in the second phase that is based on the test rig method, through which is measured the thermal conductivity of pebble bed material. In this framework, the experimental procedure and the measured results obtained varying the temperature, are presented and discussed.

  10. Packed fluidization, enhancement of heat transfer in pebble bed and thermonuclear fusion technology

    International Nuclear Information System (INIS)

    Packed fluidization is a technique in which small particles (size: 100-800 μm) are allowed to fluidize in the interstices of stationary pebbles (size: >1.0 mm). Packed fluidization enhances the rate of heat transfer in pebble bed at low operative gas velocity as well as at low pressure drop across the bed. Experiments were conducted to study heat transfer in unary packed bed and binary packed fluidized bed using lithium titanate and alumina pebbles (size: 3-10 mm) and lithium titanate and silica particles (size: 231-780 μm). It was found that due to packed fluidization the rate of heat transfer is enhanced and arms of the effective thermal conductivity this enhancement was up to 260%. Low thermal conductivity of pebble bed of solid breeder materials is one of the adverse key issues which must be addressed properly for the successful development of the thermonuclear fusion technology. Packed fluidization enhances the effective thermal conductivity of the pebble bed of solid breeder materials in the Test Blanket Module (TBM) of ITER type fusion reactor. (author)

  11. Analysis of the in-pile operation and preliminary results of the post irradiation dismantling of the pebble bed assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Magielsen, A.J.; Peeters, M.M.W.; Hegeman, J.B.J.; Stijkel, M.P.; Laan, J.G. van der [NRG-Nuclear Research and Consultancy Group (Netherlands)

    2007-07-01

    The Pebble Bed Assemblies (PBA) are four tritium breeding sub scale modules, representing a segment of the European Helium Cooled Pebble Bed Test Blanket. The objective of these experiments is the study the thermomechanical behaviour of the pebble bed assemblies during irradiation. This objective will be full- filled by the analysis of changes in the in-pile temperature profiles during irradiation and the post irradiation examination of the pebble beds in the Hot Cells. The PBA has been irradiated in the HFR in Petten for 294 Full Power Days (FPD), to a dose of 2-3 dpa in Eurofer, and estimated lithium burnup of 2-3 %. Changes in the temperature profile during in-pile operation are indication for pebble bed creep compaction during first start up and the possible formation gas gaps between the pebble beds and the structure. During progressive irradiation the radial and axial differential temperatures within the breeder and beryllium pebble beds are evaluated. During start up of the sub sequent irradiation cycles (each 26 FPD) the temperature differences within the beryllium pebble beds show a slight increase suggesting changes in the structure of the pebble beds. The PBA are transported from the HFR to the Hot Cell Laboratory in upright position to maintain the gas gaps between the pebble beds and Eurofer. Various microscopy preparation techniques are used to study the deformation state of the pebble beds (signs of creep compaction and sintering), formation of gas gaps between the pebble beds and structural materials and the interaction layers between eurofer-ceramic and eurofer-beryllium. In this paper first results on the Post Irradiation examination are given. (orig.)

  12. Preparation and characterization of Li4SiO4 ceramic pebbles by graphite bed method

    International Nuclear Information System (INIS)

    Highlights: • Lithium orthosilicate pebbles were fabricated by a new graphite bed process. • Two routes using different raw materials have been conducted in this work. • The fabricated pebbles exhibit a high relative density with uniform microstructure. • This method is short and simple as the pebbles could be fabricated in a continuous process. - Abstract: Lithium-based ceramics have long been recognized as tritium breeding materials in fusion reactor blankets. Lithium orthosilicate (Li4SiO4) is one of these materials and has been recommended by many ITER research teams as the first selection for the solid tritium breeder. In this paper, the fabrication of Li4SiO4 pebbles used as tritium breeder by a graphite bed method was studied for the first time. Ceramic powders and deionized water were mixed and ball milled to obtain homogeneous suspensions. And then the ceramic suspensions were dispersed on spread graphite powder through nozzles. Spherical droplets with highly uniform size were formed by the surface tension of the liquid droplets. The droplets converted into green pebbles after drying. After calcination and sintering, Li4SiO4 pebbles with desired size and shape were prepared. The obtained Li4SiO4 pebbles had narrow size distribution and favorable sphericity. Thermal analysis, phase analysis and microstructure observation of the pebbles were carried out systematically. Properties of the prepared pebbles were also characterized for crushing load strength, density and porosity, etc. The values were found to be conforming to the desired properties for used as solid breeder

  13. Granular Dynamics in Pebble Bed Reactor Cores

    Science.gov (United States)

    Laufer, Michael Robert

    This study focused on developing a better understanding of granular dynamics in pebble bed reactor cores through experimental work and computer simulations. The work completed includes analysis of pebble motion data from three scaled experiments based on the annular core of the Pebble Bed Fluoride Salt-Cooled High- Temperature Reactor (PB-FHR). The experiments are accompanied by the development of a new discrete element simulation code, GRECO, which is designed to offer a simple user interface and simplified two-dimensional system that can be used for iterative purposes in the preliminary phases of core design. The results of this study are focused on the PB-FHR, but can easily be extended for gas-cooled reactor designs. Experimental results are presented for three Pebble Recirculation Experiments (PREX). PREX 2 and 3.0 are conventional gravity-dominated granular systems based on the annular PB-FHR core design for a 900 MWth commercial prototype plant and a 16 MWth test reactor, respectively. Detailed results are presented for the pebble velocity field, mixing at the radial zone interfaces, and pebble residence times. A new Monte Carlo algorithm was developed to study the residence time distributions of pebbles in different radial zones. These dry experiments demonstrated the basic viability of radial pebble zoning in cores with diverging geometry before pebbles reach the active core. Results are also presented from PREX 3.1, a scaled facility that uses simulant materials to evaluate the impact of coupled fluid drag forces on the granular dynamics in the PB-FHR core. PREX 3.1 was used to collect first of a kind pebble motion data in a multidimensional porous media flow field. Pebble motion data were collected for a range of axial and cross fluid flow configurations where the drag forces range from half the buoyancy force up to ten times greater than the buoyancy force. Detailed analysis is presented for the pebble velocity field, mixing behavior, and residence time

  14. Drucker-Prager-Cap creep modelling of pebble beds in fusion blankets

    International Nuclear Information System (INIS)

    Modelling of thermal and mechanical behaviour of pebble beds for fusion blankets is an important issue to understand the interaction of solid breeder and beryllium pebble beds with the surrounding structural material. Especially the differing coefficients of thermal expansion of these materials cause high stresses and strains during irradiation induced volumetric heating. To describe this process, the coupled thermomechanical behaviour of both pebble bed materials has to be modelled. Additionally, creep has to be considered contributing to bed deformations and stress relaxation. Motivated by experiments, we use a continuum mechanical approach called Drucker-Prager/Cap theory to model the macroscopic pebble bed behaviour. The model accounts for pressure dependent shear failure, inelastic hardening, and volumetric creep. The elastic part is described by a nonlinear elasticity law. The model has been implemented by user-defined routines in the commercial finite-element code ABAQUS. To check the numerics, the implementation is compared to an analytical solution. Furthermore, the Drucker-Prager/Cap tool is applied to a single ceramic breeder bed subject to creep under volumetric heating

  15. Development of fabrication technologies for advanced tritium breeder pebbles by the sol–gel method

    International Nuclear Information System (INIS)

    Highlights: • Li2TiO3 with excess Li (Li2+xTiO3+y) was developed as an advanced tritium breeder. • Pebble fabrication by the sol–gel method is a promising technique for the mass production of advanced tritium breeder pebbles. • To increase the density of the sintered Li2+xTiO3+y pebbles, the sintering temperature was changed. • At 1353 K, the density of the pebbles increased to approximately 85% T.D. without any increase in the grain size. -- Abstract: Demonstration power plant (DEMO) reactors require advanced tritium breeders with high thermal stability. Li2TiO3 with excess Li (Li2+xTiO3+y) was developed as an advanced tritium breeder. Pebble fabrication by the sol–gel method is a promising technique for the mass production of advanced tritium breeder pebbles. I have been developing a sol–gel technique for fabricating Li2+xTiO3+y pebbles, and the next step is to optimize the granulation conditions to reach the target value. In a previous study, the average grain size on the surfaces and cross sections of sintered Li2+xTiO3+y pebbles was large whereas the theoretical density (T.D.) of these pebbles was small. To increase the density of the sintered Li2+xTiO3+y pebbles, the sintering temperature was changed, and at 1353 K, the density of the pebbles increased to approximately 85% T.D. without any increase in the grain size. This suggests that the pore size in the sintered Li2+xTiO3+y pebbles decreased because of sintering in vacuum and argon

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

    International Nuclear Information System (INIS)

    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

  17. 3D tomography analysis of the inner structure of pebbles and pebble beds

    International Nuclear Information System (INIS)

    An analytical tool to monitor the arrangement of pebbles in a pebble bed as well as the morphology of gas bubbles in as fabricated and neutron irradiated beryllium pebbles is presented. The context of this study is the Helium Cooled Pebble Bed (HPCB) blanket design for the forthcoming generation of fusion reactors. The thermal-mechanical behavior of pebble beds is a basic issue for the HPCB. It depends strongly on the configuration of the pebbles in the bed, and in particular on the number of contacts between pebbles, and between pebbles and the blanket walls. The related contact surfaces play also a major role. The knowledge on the inner structure of the pebbles is required since during the life cycle of a power reactor helium and tritium bubbles are produced inside the beryllium pebbles and the tritium build-up can be in excess of several kilograms, being thereby a key safety issue. All the non-destructive analyses are based on 3D computer aided microtomography using a very powerful synchrotron radiation x-ray source with high spatial resolution. The data analysis relies on a topological operator called filtered medial line applied to the entire data volumes and the related graph representation. By this technique the number of contacts between the pebbles in pebble packs and their angular distribution are obtained, as well as the corresponding contact surfaces. The evaluation of bubble sizes and densities in single pebbles, the assessment of the pore channel network topology, the 3D reconstruction of the fraction of interconnected bubble porosity, and the open-to-closed-porosity ratio are among the most interesting findings. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  18. Survey of dust production in pebble bed reactor cores

    International Nuclear Information System (INIS)

    Highlights: → We review potential sources of the graphite dust found in the German pebble bed reactors. → Available literature on graphite wear coefficients in pebble bed core-like conditions is reviewed. → Limited conclusions and remaining open questions are discussed. - Abstract: Graphite dust produced via mechanical wear from the pebbles in a pebble bed reactor is an area of concern for licensing. Both the German pebble bed reactors produced graphite dust that contained activated elements. These activation products constitute an additional source term of radiation and must be taken under consideration during the conduct of accident analysis of the design. This paper discusses the available literature on graphite dust production and measurements in pebble bed reactors. Limited data is available on the graphite dust produced from the AVR and THTR-300 pebble bed reactors. Experiments that have been performed on wear of graphite in pebble-bed-like conditions are reviewed. The calculation of contact forces, which are a key driving mechanism for dust in the reactor, are also included. In addition, prior graphite dust predictions are examined, and future areas of research are identified.

  19. A Pebble Bed Reactor cross section methodology

    International Nuclear Information System (INIS)

    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.

  20. Mechanism analysis of quasi-static dense pebble flow in pebble bed reactor using phenomenological approach

    International Nuclear Information System (INIS)

    Highlights: ► We introduced four basic forms of phenomenological method for pebble flow. ► We discussed the physical nature of the quasi-static pebble flow. ► We verified the applicability of the discrete element method. ► We investigated the parameter effects on quasi-static pebble flow. - Abstract: By means of the four basic forms of the phenomenological method, experimental results have intuitionally disclosed the physical mechanism from various views of the quasi-static pebble flow in a pebble bed reactor and successfully verified the availability of the discrete element method, on which the parameter effects have been investigated, including different base cone angle and different friction coefficient. The flow fields under different parameters have been discussed. On the basis of these researches, a framework of the general understanding of pebble flow mechanism has been drawn; many essential problems are discussed, including the interpretation of the quasi-static pebble flow, force analysis inside the pebble packing, propagation and distribution of the voids, internal equilibrium arches, competition mechanism, internal collapse, self-organization, equivalent shear force, equivalent normal force, the physical process of stagnant zone's influence on the overall flow field, and so on. All of these are very helpful to understand the physical mechanism of the quasi-static pebble flow in a pebble bed reactor.

  1. Uncertainty and sensitivity analysis of filling fraction of pebble bed in pebble bed HTR

    International Nuclear Information System (INIS)

    Highlights: • An analysis approach is proposed to conduct SAU analysis of filling fraction of pebble bed core. • The contribution of uncertainty in filling fraction to key parameters of pebble bed core is quantified. • The primary drivers of the uncertainty in the key parameters are identified by sensitivity analysis. • Mechanism of effect of uncertainty in the filling fraction to key parameters is studied in depth. - Abstract: The filling fraction of pebble bed in each small region has some uncertainty, which will contribute to the total uncertainty in the key parameters of pebble bed core, such as power peak, axial offset, keff and so on. In fact, the heavy metal content and graphite content of the corresponding region will change synchronously due to the perturbation of filling fraction but the ratio of atomic number density of moderator to the fuel (carbon–uranium) is still constant. To investigate these effects, the Chinese demonstration plant HTR-PM was selected as the research object and the VSOP code and CUSA package were used to conduct detailed analysis of the influence of the uncertainty in the filling fraction on the HTR-PM output variables of interest, based on the propagation of input uncertainties by using statistical sampling method to calculate uncertainty and sensitivity information from the simulation results. At the same time, uncertainty and sensitivity analysis of uranium loading had also been conducted for comparative analysis to study the mechanism of effect of uncertainty in the filling fraction to key parameters, and therefore only the heavy metal content of the corresponding region changes in the presence of perturbation of uranium loading. Finally, the propagated uncertainty in the power peak, axial offset and keff of HTR-PM core was obtained and the primary drivers of the uncertainty in the key parameters were identified by sensitivity analysis

  2. Experimental study and analysis of the purge gas pressure drop across the pebble beds for the fusion HCPB blanket

    International Nuclear Information System (INIS)

    Highlights: ► The pressure drop significantly increases with decreasing the pebbles diameter. ► The pressure drop slightly increases with increasing the packing factor. ► The pressure drop is directly proportional to pebble bed length and inlet pressure. ► Predictions of Ergun equation agree well with the measured values of pressure drop. ► The filters resistance has a small contribution to the total pressure drop. -- Abstract: The lithium ceramic and beryllium pebble beds of the breeder units (BU), in the fusion breeding blanket, are purged by helium to extract the bred tritium. Therefore, the objective of this study is to support the design of the BU purge gas system by studying the effect of pebbles diameter, packing factor, pebble bed length, and flow inlet pressure on the purge gas pressure drop. The pebble bed was formed by packing glass pebbles in a rectangular container (56 mm × 206 mm × 396 mm) and was integrated into a gas loop to be purged by helium at BU-relevant pressures (1.1–3.8 bar). To determine the pressure drop across the pebble bed, the static pressure was measured at four locations along the pebble bed as well as at the inlet and outlet locations. The results show: (i) the pressure drop significantly increases with decreasing the pebbles diameter and slightly increases with increasing the packing factor, (ii) for a constant inlet flow velocity, the pressure drop is directly proportional to the pebble bed length and inlet pressure, and (iii) predictions of Ergun's equation agree well with the experimental values of the pressure drop

  3. Trial examination of direct pebble fabrication for advanced tritium breeders by the emulsion method

    Energy Technology Data Exchange (ETDEWEB)

    Hoshino, Tsuyoshi, E-mail: hoshino.tsuyoshi@jaea.go.jp

    2014-10-15

    Highlights: • The integration of raw material preparation and granulation is proposed as a new direct pebble fabrication process. • The emulsion method granulates gel spheres of Li{sub 2}CO{sub 3} and TiO{sub 2} or SiO{sub 2}. • The gel spheres are calcined and sintered in air. • The crush load of the sintered Li{sub 2}TiO{sub 3} or Li{sub 4}SiO{sub 4} pebbles obtained is 37.2 or 59.3 N, respectively. - Abstract: Demonstration power plant reactors require advanced tritium breeders with high thermal stability. For the mass production of advanced tritium breeder pebbles, pebble fabrication by the emulsion method is a promising technique. To develop the most efficient pebble fabrication method, a new direct pebble fabrication process utilizing the emulsion method was implemented. A prior pebble fabrication process consisted of the preparation of raw materials followed by granulation. The new process integrates the preparation and granulation of raw materials. The slurry for the emulsion granulation of Li{sub 2}TiO{sub 3} or Li{sub 4}SiO{sub 4} as a tritium breeder consists of mixtures of Li{sub 2}CO{sub 3} and TiO{sub 2} or SiO{sub 2} at specific ratios. Subsequently, gel spheres of tritium breeders are fabricated by controlling the relative flow speeds of slurry and oil. The average diameter and crush load of the obtained sintered Li{sub 2}TiO{sub 3} or Li{sub 4}SiO{sub 4} pebbles were 1.0 or 1.5 mm and 37.2 or 59.3 N, respectively. The trial fabrication results suggest that the new process has the potential to increase the fabrication efficiency of advanced tritium breeder pebbles.

  4. Trial examination of direct pebble fabrication for advanced tritium breeders by the emulsion method

    International Nuclear Information System (INIS)

    Highlights: • The integration of raw material preparation and granulation is proposed as a new direct pebble fabrication process. • The emulsion method granulates gel spheres of Li2CO3 and TiO2 or SiO2. • The gel spheres are calcined and sintered in air. • The crush load of the sintered Li2TiO3 or Li4SiO4 pebbles obtained is 37.2 or 59.3 N, respectively. - Abstract: Demonstration power plant reactors require advanced tritium breeders with high thermal stability. For the mass production of advanced tritium breeder pebbles, pebble fabrication by the emulsion method is a promising technique. To develop the most efficient pebble fabrication method, a new direct pebble fabrication process utilizing the emulsion method was implemented. A prior pebble fabrication process consisted of the preparation of raw materials followed by granulation. The new process integrates the preparation and granulation of raw materials. The slurry for the emulsion granulation of Li2TiO3 or Li4SiO4 as a tritium breeder consists of mixtures of Li2CO3 and TiO2 or SiO2 at specific ratios. Subsequently, gel spheres of tritium breeders are fabricated by controlling the relative flow speeds of slurry and oil. The average diameter and crush load of the obtained sintered Li2TiO3 or Li4SiO4 pebbles were 1.0 or 1.5 mm and 37.2 or 59.3 N, respectively. The trial fabrication results suggest that the new process has the potential to increase the fabrication efficiency of advanced tritium breeder pebbles

  5. 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

  6. Status of the pebble bed modular reactor

    International Nuclear Information System (INIS)

    Eskom is the South African state electricity utility, with an installed capacity of 38397 MW at the end of 1996 (some 98% of all national generating assets). It is largely coal-based with a small proportion (5%) nuclear. As part of Eskom's long-term planning process, investigations have been made into new power generation options. On reconsidering the nuclear option, Eskom identified two key issues: cost and public acceptance. It was considered that both of these were driven by the safety issues related to potential accidents and the only way to obtain competitive costs with nuclear power was to remove the potential (however remote) for accidents with significant off-site consequences. The only reactor type that was seen to meet this safety standard was the pebble bed modular reactor (PBMR). This paper discusses the PBMR project history, plant performance and design, its benefits, safety features, and current status. It concludes that the PBMR will provide South Africa with a competitive option for coastal generation and, internationally, it will be highly competitive with virtually all other generation options. (author)

  7. Preparation and characterization of Li{sub 4}SiO{sub 4} ceramic pebbles by graphite bed method

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Ming; Zhang, Yingchun, E-mail: zycustb@163.com; Xiang, Maoqiao; Liu, Zhiang

    2015-06-15

    Highlights: • Lithium orthosilicate pebbles were fabricated by a new graphite bed process. • Two routes using different raw materials have been conducted in this work. • The fabricated pebbles exhibit a high relative density with uniform microstructure. • This method is short and simple as the pebbles could be fabricated in a continuous process. - Abstract: Lithium-based ceramics have long been recognized as tritium breeding materials in fusion reactor blankets. Lithium orthosilicate (Li{sub 4}SiO{sub 4}) is one of these materials and has been recommended by many ITER research teams as the first selection for the solid tritium breeder. In this paper, the fabrication of Li{sub 4}SiO{sub 4} pebbles used as tritium breeder by a graphite bed method was studied for the first time. Ceramic powders and deionized water were mixed and ball milled to obtain homogeneous suspensions. And then the ceramic suspensions were dispersed on spread graphite powder through nozzles. Spherical droplets with highly uniform size were formed by the surface tension of the liquid droplets. The droplets converted into green pebbles after drying. After calcination and sintering, Li{sub 4}SiO{sub 4} pebbles with desired size and shape were prepared. The obtained Li{sub 4}SiO{sub 4} pebbles had narrow size distribution and favorable sphericity. Thermal analysis, phase analysis and microstructure observation of the pebbles were carried out systematically. Properties of the prepared pebbles were also characterized for crushing load strength, density and porosity, etc. The values were found to be conforming to the desired properties for used as solid breeder.

  8. Calibration of a pebble bed configuration for direct numerical simulation

    International Nuclear Information System (INIS)

    The appearance of hot spots in the pebble bed cores of High Temperature Reactors (HTR) 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. Such models need to be validated in order to gain trust in the simulation of these types of flow configurations. Direct Numerical Simulation (DNS) can serve as a reference for validation, however, it poses restrictions in terms of flow parameters and numerical tools corresponding to the available computational resources. In the present study, a wide range of numerical simulations has been performed in order to calibrate a pebble bed configuration for DNS which may serve as reference for validation. (author)

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

    International Nuclear Information System (INIS)

    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 233U 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 233U fuel from 235U feed was studied and performances of the converter and breeder reactor concepts were examined varying the key parameters

  10. Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores. [PEBBLE code

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. Effect of wall structure on pebble stagnation behavior in pebble bed reactor

    International Nuclear Information System (INIS)

    Highlights: • DEM study of wall structure role in preventing near wall crystallization is carried out. • Suggestions on pebble’s kinematic parameters and wall structure design are provided. • Triangle is better than arc and sawtooth shapes for wall structure design. • Wall structure size should be close to the scale of pebble diameter. • Suitable intervals can prevent crystallization without significantly increasing the flow resistance. - Abstract: Crystallization of pebbles in pebble bed is a crucial problem in high temperature gas-cooled pebble-bed reactors. This phenomenon usually happens along the internal surface and leads to a large number of stagnated pebbles, which poses a threat to reactor safety. In real reactor engineering, wall structures have been utilized to avoid this problem. This article verifies the crystallization phenomenon through DEM (discrete element method) simulation, and explains how wall structures work in preventing crystallization. Moreover, several kinematic parameters have been adopted to evaluate wall structures with different shapes, sizes and intervals. Detailed information shows the impact of wall structure on flow field in pebble bed. Lastly, the preferred characteristics of an effective wall structure are suggested for reactor engineering

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

    International Nuclear Information System (INIS)

    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

  13. Pebble bed reactor with a feeding device for absorber materials

    International Nuclear Information System (INIS)

    Description of a second shutdown device for pebble bed reactors with small absorber pebbles, which, if required, can be let off a storage tank and regularly trickle through a dispersion cone into the pebble bed. In the normal state the storage tank is in a low position with its outlet being obstructed by absorber pebbles filling a cylinder in which slides a piston which is firmly connected with the storage tank. By giving pressure over a line a piston in a pneumatic cylinder can be moved which lifts the storage tank. The cylinder is emptied by lifting the piston and the outlet is released. The level of the storage tank is measured by means of a probe. The whole device is installed in the prestressed concrets ceiling of the reactor. The device is proposed to be set into motion for a short moment from time to time in order to prove its operatability. (orig.)

  14. Neutronic modeling of pebble bed reactors in APOLLO2

    International Nuclear Information System (INIS)

    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 SN full-core calculation computes the entering currents in the spectrum zones subdividing the core. These currents, together with the core keff, 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 keff 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 (UO2, 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', this

  15. Optimization of MOX fuel cycles in pebble bed HTGR

    International Nuclear Information System (INIS)

    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)

  16. Control rod for a pebble bed nuclear reactor

    International Nuclear Information System (INIS)

    In order to leave the density of the pebble bed unchanged when the control rod is driven in and out, the tip of the control rod is provided with moving parts in the form of conical hemispheres or spoons. These parts move close to the control rod when it is driven in and spread out due to the effect of gravity when it is driven out. This loosens the heap of pebble shaped operating elements again. (DG)

  17. Neutronic features of pebble-bed reactors for transmutation applications

    International Nuclear Information System (INIS)

    Pebble-bed reactors offer very appealing characteristics for radioactivity confinement and for withstanding thermal transients. Besides that, pebble-bed reactors have a peculiar degree of freedom in the radius of the active core of the pebble (where the fuel is located) as compared to the outer radius of the pebble, which has a coating of pure graphite. By varying the aforementioned radius, very different types of neutron spectra can be formed, which in turn gives very different values of the average cross sections that govern the isotopic composition evolution, and particularly the elimination of the most relevant transuranics. Preliminary conclusions of this work show that there is a very broad design window for exploiting the transmutation capabilities of pebble-bed reactors in a scenario of inherent safety features. A 99,9% elimination of Pu-239 associated to a 99% elimination of Pu-240 and Pu-241 can be reached, with some increment of the Pu-242 contents (which is extremely long-lived, less radio-toxic and decays into the natural nuclide U-238). Am and Cm are also transmuted to a significant level, although some residual higher A actinides will remain. (authors)

  18. In-core fuel management for pebble-bed reactors

    Energy Technology Data Exchange (ETDEWEB)

    Milian Perez, Daniel; Rodriguez Garcia, Lorena; Garcia Hernandez, Carlos; Milian Lorenzo, Daniel, E-mail: dperez@instec.cu, E-mail: cgh@instec.cu, E-mail: dmilian@instec.cu [Higher Institute of Technologies and Applied Sciences, Havana (Cuba); Velasco, Abanades, E-mail: abanades@etsii.upm.es [Department of Simulation of Thermo Energy Systems, Polytechnic University of Madrid (Spain)

    2013-07-01

    In this paper a calculation procedure to reduce the power peak in the core of a Very High Temperature pebble bed Reactor is presented. This procedure combines the fuel depletion and the neutronic behavior of the fuel in the reactor core, modeling once-through-then-out cycles as well as cycles in which pebbles are recirculated through the core an arbitrary number of times, obtaining the asymptotic fuel-loading pattern. The procedure consists in several coupled computational codes, which are used iteratively until convergence is reached. The utilization of the MCNPX 2.6e, as one of these computational codes, is validated through the calculation of benchmarks announced by IAEA (IAEA-TECDOC-1249, 2001). To complete the verification of the calculation procedure a base case described in (Annals of Nuclear Energy 29 (2002) 1345-1364), was performed. The procedure has been applied to a model of Pebble Bed Modular Reactor (200 MW) design. (author)

  19. Thermal behaviour and tritium management for in-pile testing of the pebble bed assemblies in the HFR in Petten

    International Nuclear Information System (INIS)

    Four pebble-bed assemblies are to be irradiated in the HFR in Petten with the objective to study the thermo-mechanical behaviour of the breeder ceramic pebble beds during irradiation. The thermo-mechanical behaviour of the pebble bed assemblies was calculated in a 2D axi-symmetric model in MARC. In this approach there could not be accounted for the influence of thermocouple tubes on the temperature distribution in the assembly, because these are distributed in the assembly in a non axi-symmetric manner. The solution for this problem was to expand the model to a 3D model used for thermal computations only. For safety reasons the tritium production in the breeder and permeation through the first and second containment must be estimated before the in-pile experimentation begins. In order to do so, the calculated thermal distribution is used as input for the enhanced two-dimensional finite element model in MARC. Adaptations are made in the 2D model by adding the capability of performing mass flux calculations. This paper describes the finite element models used for computation of the temperature distribution and the tritium flux through the pebble bed assembly. The results of these calculations are critical for a safety assessment of the in-pile operation of the experiment and will give a better understanding of the in-pile behaviour on temperature and tritium management in advance. (orig.)

  20. A study on evaluation of pebble flow velocity with modification of the kinematic model for pebble bed reactor

    International Nuclear Information System (INIS)

    Highlights: ► A modified kinematic method is proposed for analysis of pebble flow velocity. ► Experiments are performed to derive the coefficients and to verify the results. ► The method and result can be used for the advanced analysis of pebble bed reactor. - Abstract: A pebble bed reactor is filled by a large number of pebbles, which are randomly piled up in the core region. During the process of fuel loading and extraction, the pebbles flow downward through the core. The basic physics of the dense granular flow such as pebble flow is not fully understood; hence, the dynamic core of the pebble bed reactor has been a subject of concern among designers and regulators. The kinematic model is one of the representative models for the reconstruction of the granular flow velocity, however, it is noted that there are some limitations in the reconstruction ability. In this study, a modified kinematic model was proposed to enhance the reconstruction ability of the pebble velocity profile. Pebble flow experiments were performed to derive the coefficients needed for the modified kinematic model and to verify the reconstruction ability and the applicability of the proposed method in the annular core. The modified kinematic model can contribute to accurate velocity evaluation as well as large applicability for the specific core types such as an annular core. Also, the results can be used for reference data in the design of a pebble bed reactor

  1. Pebble Bed Reactor review update. Fiscal year 1979 annual report

    International Nuclear Information System (INIS)

    Updated information is presented on the Pebble Bed Reactor (PBR) concept being developed in the Federal Republic of Germany for electricity generation and process heat applications. Information is presented concerning nuclear analysis and core performance, fuel cycle evaluation, reactor internals, and safety and availability

  2. Pebble Bed Reactor review update. Fiscal year 1979 annual report

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    Updated information is presented on the Pebble Bed Reactor (PBR) concept being developed in the Federal Republic of Germany for electricity generation and process heat applications. Information is presented concerning nuclear analysis and core performance, fuel cycle evaluation, reactor internals, and safety and availability.

  3. PEBBLE: a two-dimensional steady-state pebble bed reactor thermal hydraulics code

    International Nuclear Information System (INIS)

    This report documents the local implementation of the PEBBLE code to treat the two-dimensional steady-state pebble bed reactor thermal hydraulics problem. This code is implemented as a module of a computation system used for reactor core history calculations. Given power density data, the geometric description in (RZ), and basic heat removal conditions and thermal properties, the coolant properties, flow conditions, and temperature distributions in the pebble fuel elements are predicted. The calculation is oriented to the continuous fueling, steady state condition with consideration of the effect of the high energy neutron flux exposure and temperature history on the thermal conductivity. The coolant flow conditions are calculated for the same geometry as used in the neutronics calculation, power density and fluence data being used directly, and temperature results are made available for subsequent use

  4. PEBBLE: a two-dimensional steady-state pebble bed reactor thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1981-09-01

    This report documents the local implementation of the PEBBLE code to treat the two-dimensional steady-state pebble bed reactor thermal hydraulics problem. This code is implemented as a module of a computation system used for reactor core history calculations. Given power density data, the geometric description in (RZ), and basic heat removal conditions and thermal properties, the coolant properties, flow conditions, and temperature distributions in the pebble fuel elements are predicted. The calculation is oriented to the continuous fueling, steady state condition with consideration of the effect of the high energy neutron flux exposure and temperature history on the thermal conductivity. The coolant flow conditions are calculated for the same geometry as used in the neutronics calculation, power density and fluence data being used directly, and temperature results are made available for subsequent use.

  5. PEBBED ANALYSIS OF HOT SPOTS IN PEBBLE-BED REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Abderrafi M. Ougouag; Hans D. Gougar; William K. Terry; Frederik Reitsma; Wessel Joubert

    2005-09-01

    The Idaho National Laboratory’s PEBBED code and simple probability considerations are used to estimate the likelihood and consequences of the accumulation of highly reactive pebbles in the region of peak power in a pebble-bed reactor. The PEBBED code is briefly described, and the logic of the probability calculations is presented in detail. The results of the calculations appear to show that hot-spot formation produces only moderate increases in peak accident temperatures, and no increases at all in normal operating temperatures.

  6. Computational prediction of dust production in pebble bed reactors

    International Nuclear Information System (INIS)

    Highlights: ► Finite element analysis of frictional contact. ► Plasticity taken into account for nuclear graphite at room temperature. ► Prediction of order of magnitude for dust loading in PBRs. ► Archard wear model for wear mass calculations. - Abstract: This paper describes the computational modeling and simulation of graphite pebbles in frictional contacts as anticipated in a pebble bed reactor. For the high temperature gas-cooled reactor, the potential dust generation from frictional contact at the surface of pebbles and the subsequent lift-off and transport of dust and absorbed fission products are of safety concern at elevated temperatures under an air ingress accident. The aim of this work is to perform a computational study to estimate the quantity of the nuclear grade graphite dust produces from a typical anticipated configuration.

  7. Regulation of the pebble flow in a pebble-bed reactor

    International Nuclear Information System (INIS)

    The cylindrical core tank of the pebble bed reactor has a funnel-shaped outlet at its bottom end with one single discharge tube. The cylindrical part of the pebble bed has a height-to-diameter ratio of 0.5. In order to achieve an approximately constant vertical velocity of the OTTO program in the Juelich AVR and the Uentrop THTR reactors, respectively, there is a small cone mounted on supports in the outlet. Its tip is pointed upward and its largest diameter has a distance to the wall of the funnel sufficient to accommodate the diameter of the spheres. At high powers the outlet and the cone are built up of graphite blocks. They are equipped with numerous vertical channels for coolant passage. The cone may also be a rotationally symmetrical body with a rhombic longitudinal section which rests upon the outlet on radial bars. (orig./PW)

  8. Phenomenological method investigation of pebble flow dynamics in two-dimensional two-region pebble-bed reactor

    International Nuclear Information System (INIS)

    By means of the four basic forms of the phenomenological method, the experimental research was carried out according to the principle of similarity criterion to simulate 2D pebble flow dynamics of high-temperature gas-cooled reactor. The result indicates that the test with circulating pebble-loading mode obviously presents better the situation in the real pebble bed reactor. The pebble flow dynamics spreads from bottom to top and from middle to sides. The movement of pebbles in central region is faster than that in annulus region and has no laminar characteristics performance. The mixed zone exists between central region and annulus region, and the distinct stagnant zone also exists at pebble bed bottom corner. (authors)

  9. Thorium and plutonium utilisation in pebble-bed modular reactor

    International Nuclear Information System (INIS)

    Thorium and plutonium utilisation in a high temperature gas-cooled pebble-bed reactor is investigated with the aim to predict the economic value of vast thorium reserves in Turkey. A pebble-bed reactor of the type designed by PBMR Pty. of South Africa is taken as the investigated system. The equilibrium core of a PBMR is considered and neutronics analyses of such a core are performed through the use of the SCALE-4.4 computer code system KENOV.a module. Various cross-section libraries are used to calculate the criticality of the core. Burn-up calculations of the core are performed by coupling the KENOV.a module with the ORIGEN-S module. Calculations are carried out for various U-Th, U-Pu-Th and U-Pu combinations. The results are preliminary in nature and the work is currently proceeding as planned. (author)

  10. Numerical simulation on pebble dynamics of two-dimensional two-region pebble-bed reactor using phenomenological method

    International Nuclear Information System (INIS)

    Discrete element method was used to simulate the pebble dynamics in the high-temperature gas-cooled reactor core, based on the two-dimensional pebble dynamics experiments. Phenomenological method was used to analyze the formation of the two-region distribution, the central, mixing and stagnant regions, and the velocity distribution. The simulation results show that a stable central region is formed, mixing zones between the central and annular regions and stagnant regions are observed in current simulation. The closer to the bottom of the pebble bed, the more uneven the vertical pebble velocity and the bigger the horizontal diffusion. (authors)

  11. Modeling of laminar forced convection in spherical- pebble packed beds

    International Nuclear Information System (INIS)

    There are many parameters that have significant effects on forced convection heat transfer in packed beds, including Reynolds and Prandtl numbers of flow, porosity, pebble geometry, local flow conditions, wall and end effects. In addition, there have been many experimental investigations on forced convection heat transfer in packed beds and each have studied the effect of some of these parameters. Yet, there is not a reliable correlation that includes the effect of main parameters: at the same time, the prediction of precise correct limits for very low and high Reynolds numbers is off hand. In this article a general well-known model of convection heat transfer from isothermal bodies, next to some previous reliable experimental data has been used as a basis for a more comprehensive and accurate correlation to calculate the laminar constant temperature pebble-fluid forced convection heat transfer in a homogeneous saturated bed with spherical pebbles. Finally, for corroboration, the present results are compared with previous works and show a very good agreement for laminar flows at any Prandtl number and all porosities

  12. Transmutation of plutonium in pebble bed type high temperature reactors

    International Nuclear Information System (INIS)

    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/GWea is achieved. The discharges of plutonium and minor actinides are then 450 and 110 kg/GWea, respectively. (author)

  13. Modeling of realistic pebble bed reactor geometries using the Serpent Monte Carlo code

    International Nuclear Information System (INIS)

    Highlights: • The explicit stochastic geometry model in Serpent is documented. • A pebble bed criticality benchmark was calculated demonstrating the geometry model. • Stochastic pebble configurations were obtained from discrete element simulations. • Results deviate from experiments but are in line with example calculations. - Abstract: This paper documents the models available in Serpent for high temperature reactor (HTR) calculations. It is supplemented by a calculation example of ASTRA critical pebble bed experiments. In the pebble bed reactor modeling, different methods have been used to model the double heterogeneity problem occurring in pebble bed reactor calculations. A solution was sought to avoid unphysical simplifications in the pebble bed modeling and the stochastic geometry modeling features available in the Monte Carlo code Serpent were applied for exact placement of pebbles and fuel particles. Randomly packed pebble beds were produced in discrete element method (DEM) simulations and fuel particles were positioned randomly inside the pebbles. Pebbles and particles are located using a Cartesian search mesh, which provides necessary computational efficiency. Serpent uses Woodcock delta-tracking which provides efficient neutron tracking in the complicated geometries. This detailed pebble bed modeling approach was tested by calculating the ASTRA criticality benchmark experiment done at the Kurchatov Institute in 2004. The calculation results are in line with the sample calculations provided with the benchmark documentation. The material library selected for the calculations has a major effect on the results. The difference in graphite absorption cross section is considered the cause of this result. The model added in Serpent is very efficient with a calculation time slightly higher than with a regular lattice approximation. It is demonstrated that Serpent can be used for pebble bed reactor calculations with minimal geometric approximations as it

  14. Studies on air ingress for pebble bed reactors

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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 235U enrichment of the fuel, impurities in the moderator pebbles, and the density and impurity content of the radial reflector. Calculations of keff 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 keff 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

  16. Requirements for helium cooled pebble bed blanket and R and D activities

    Energy Technology Data Exchange (ETDEWEB)

    Carloni, D., E-mail: dario.carloni@kit.edu; Boccaccini, L.V.; Franza, F.; Kecskes, S.

    2014-10-15

    This work aims to give an outline of the design requirements of the helium cooled pebble bed (HCPB) blanket and its associated R and D activities. In DEMO fusion reactor the plasma facing components have to fulfill several requirements dictated by safety and process sustainability criteria. In particular the blanket of a fusion reactor shall transfer the heat load coming from the plasma to the cooling system and also provide tritium breeding for the fuel cycle of the machine. KIT has been investigating and developed a helium-cooled blanket for more than three decades: the concept is based on the adoption of separated small lithium orthosilicate (tritium breeder) and beryllium (neutron multiplier) pebble beds, i.e. the HCPB blanket. One of the test blanket modules of ITER will be a HCPB type, aiming to demonstrate the soundness of the concept for the exploitation in future fusion power plants. A discussion is reported also on the development of the design criteria for the blanket to meet the requirements, such as tritium environmental release, also with reference to the TBM. The selection of materials and components to be used in a unique environment as the Tokamak of a fusion reactor requires dedicated several R and D activities. For instance, the performance of the coolant and the tritium self-sufficiency are key elements for the realization of the HCPB concept. Experimental campaigns have been conducted to select the materials to be used inside the solid breeder blanket and R and D activities have been carried out to support the design. The paper discusses also the program of future developments for the realization of the HCPB concept, also focusing to the specific campaigns necessary to qualify the TBM for its implementation in the ITER machine.

  17. Requirements for helium cooled pebble bed blanket and R and D activities

    International Nuclear Information System (INIS)

    This work aims to give an outline of the design requirements of the helium cooled pebble bed (HCPB) blanket and its associated R and D activities. In DEMO fusion reactor the plasma facing components have to fulfill several requirements dictated by safety and process sustainability criteria. In particular the blanket of a fusion reactor shall transfer the heat load coming from the plasma to the cooling system and also provide tritium breeding for the fuel cycle of the machine. KIT has been investigating and developed a helium-cooled blanket for more than three decades: the concept is based on the adoption of separated small lithium orthosilicate (tritium breeder) and beryllium (neutron multiplier) pebble beds, i.e. the HCPB blanket. One of the test blanket modules of ITER will be a HCPB type, aiming to demonstrate the soundness of the concept for the exploitation in future fusion power plants. A discussion is reported also on the development of the design criteria for the blanket to meet the requirements, such as tritium environmental release, also with reference to the TBM. The selection of materials and components to be used in a unique environment as the Tokamak of a fusion reactor requires dedicated several R and D activities. For instance, the performance of the coolant and the tritium self-sufficiency are key elements for the realization of the HCPB concept. Experimental campaigns have been conducted to select the materials to be used inside the solid breeder blanket and R and D activities have been carried out to support the design. The paper discusses also the program of future developments for the realization of the HCPB concept, also focusing to the specific campaigns necessary to qualify the TBM for its implementation in the ITER machine

  18. Integrated design approach of the pebble bed modular using models

    International Nuclear Information System (INIS)

    The Pebble Bed Modular Reactor (PBMR) is the first pebble bed reactor that will be utilised in a high temperature direct Brayton cycle configuration. This implies that there are a number of unique features in the PBMR that extend from the German experience base. One of the challenges in the design of the PBMR is managing the integrated design process between the designers, the physicists and the analysts. This integrated design process is managed through model-based development work. Three-dimensional CAD models are constructed of the components and parts in the reactor. From the CAD models, CFD models, neutronic models, shielding models, FEM models and other thermodynamic models are derived. These models range from very simple models to extremely detailed and complex models. The models are used in legacy software as well as commercial off-the-shelf software. The different models are also used in code-to-code comparisons to verify the results. This paper will briefly discuss the different models and the interaction between the models, showing the iterative design process that is used in the development of the reactor at PBMR. (author)

  19. Steady-state thermal-hydraulic of pebble bed blanket on hybrid reactor

    International Nuclear Information System (INIS)

    This paper gives thermal-hydraulic studies of pebble bed blanket on Hybrid Reactor. The concept of whole pebble bed blanket and the cooling methods are presented. The thermal-hydraulic characteristics of pebble bed are summarized. The theoretical model and code for solving heat transfer and flowing are presented. By using this code the calculation and analysis of thermal hydraulic of pebble bed Blanket of Hybrid Reactor are also given. In order to improve the flexibility, safety and economy, the authors select pebble beds not only to breed Tritium, but also to breed fission material and to multiply neutron. 5 MPa Helium is used as coolant and 0.05 MPa-0.1 MPa Helium is used as Purge gas. The heat transfer mechanisms of pebble bed are very complicated which include conduction, convection and radiation. In order to study the thermal-hydraulic of the bed, the authors just simply consider it as homogeneous and continuous binary phase medium as that used in the porous medium at the condition that the size of the bed is much greater than that of the balls. The coolant or the purge gas flowing through the bed is just considered existing a cooling source in the bed. It also significantly influences the effective conductivity's of the bed. Porous fraction, the main factor of the bed depends on the geometry position and parameters. From this model, one can obtain the thermal-hydraulic governing equations of the bed

  20. Gas Reactor International Cooperative Program. Interim report. Safety and licensing evaluaion of German Pebble Bed Reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    1978-09-01

    The Pebble Bed Gas Cooled Reactor, as developed in the Federal Republic of Germany, was reviewed from a United States Safety and Licensing perspective. The primary concepts considered were the steam cycle electric generating pebble bed (HTR-K) and the process heat pebble bed (PNP), although generic consideration of the direct cycle gas turbine pebble bed (HHT) was included. The study examines potential U.S. licensing issues and offers some suggestions as to required development areas.

  1. Deleterious Thermal Effects Due To Randomized Flow Paths in Pebble Bed, and Particle Bed Style Reactors

    Science.gov (United States)

    Moran, Robert P.

    2013-01-01

    A review of literature associated with Pebble Bed and Particle Bed reactor core research has revealed a systemic problem inherent to reactor core concepts which utilize randomized rather than structured coolant channel flow paths. For both the Pebble Bed and Particle Bed Reactor designs; case studies reveal that for indeterminate reasons, regions within the core would suffer from excessive heating leading to thermal runaway and localized fuel melting. A thermal Computational Fluid Dynamics model was utilized to verify that In both the Pebble Bed and Particle Bed Reactor concepts randomized coolant channel pathways combined with localized high temperature regions would work together to resist the flow of coolant diverting it away from where it is needed the most to cooler less resistive pathways where it is needed the least. In other words given the choice via randomized coolant pathways the reactor coolant will take the path of least resistance, and hot zones offer the highest resistance. Having identified the relationship between randomized coolant channel pathways and localized fuel melting it is now safe to assume that other reactor concepts that utilize randomized coolant pathways such as the foam core reactor are also susceptible to this phenomenon.

  2. Dynamics of a small direct cycle pebble bed HTR

    International Nuclear Information System (INIS)

    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)

  3. Design windows for accelerator driven pebble-bed transmutators

    International Nuclear Information System (INIS)

    Nuclear waste transmutation can be achieved by different strategies. In this paper, the studies are focused in the 'Once Through' scenario, consisting in the nuclear waste transmutation until a maximum burnup (BU) is achieved. After transmutation, the fuel elements can be disposed in a Deep Storage Facility (DSF.) The main advantage of this strategy is that only one reprocess step is necessary. The drawback of this strategy consists mainly in the need of a fuel element design capable of withstanding very high burn-ups. It has been demonstrated that pebbles fuel elements in a pebble bed reactor design can withstand 700 MWd/Kg BU. This reactor presents the possibility of attainment different neutron spectrum with different fuel element designs, presents good safety characteristics, and the possibility of replacing the fuel elements easily inside the reactor (necessary for recycling strategies.) The transmutation process can be achieved in two steps. The first one, as a critical reactor, and the second one, as a subcritical assembly driven by an accelerator. In this paper, the optimum spectrum for the 'Once Through' strategy is presented, and some safety characteristics of the subcritical assembly are introduced. (authors)

  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. Pebble bed reactor with one-zone core

    International Nuclear Information System (INIS)

    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. (orig./PW)

  6. Neutron wave experiment in a graphite pebble-bed system

    International Nuclear Information System (INIS)

    The propagation of neutron waves through a Type-AVR graphite pebble-bed is studied. Use of a sinusoidally modulated source of neutrons is equivalent to 'poisoning' a moderator with a 1/v poison. The inverse relaxation length of the neutron wave amplitude and the variation of the phase angle as function of position are dependent upon the frequency of modulation and the neutron diffusion and thermalization parameters of the media in which the waves are being propagated. The diffusion coefficient D0 of a system of graphite spheres is determined to a high accuracy. In the termal energy range a streaming correction of 14,8% is necessary if for calculation the graphite of the spheres is homogenized. (orig.)

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

    International Nuclear Information System (INIS)

    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.)

  8. Renewable lower reflector structure for a high temperature pebble bed reactor

    International Nuclear Information System (INIS)

    The description is given of a renewable lower reflector structure for a high temperature pebble bed reactor of the type comprising a cylindrical or prismatic graphite vessel wrapped in concrete and terminating at its lower end with a conical or pyramidal bottom fitted with a central aperture allowing the pebbles to be discharged by gravity. This structure includes a bed of several layers of protective graphite pebbles on the bottom and, fitted vertically so as to be removable along the centre line of the central aperture through the reflector and the concrete, a graphite block drilled in its centre to allow the discharge of the fuel pebbles and the protective pebbles. The graphite block rises above the level of the central aperture by an extent corresponding to the thickness of the bed when the reactor is working

  9. Experimental studies on heat transfer and pressure drop in pebble bed test facility

    International Nuclear Information System (INIS)

    Indian program for development of high temperature reactor and its utilization to supply process heat aimed to develop alternate fuel carrier to substitute petroleum based transport fuel, which has very small reserves in India and results in large import bills. Hydrogen is an attractive energy carrier for transport applications. It can be produced by splitting water which requires either electricity or process heat at high temperatures or both depending upon the process selected. BARC is carrying out design of a 600 MWth reactor capable of supplying process heat at around 1000 °C as required for hydrogen production. For this reactor various design options with respect to fuel configurations, such as prismatic bed and pebble bed were considered for thermal hydraulics analysis. Coolant options such as molten lead and molten salt were analyzed. Studies carried out indicate selection of pebble bed reactor core with molten salt as coolant. Thermal-hydraulic studies are required for pebble bed reactor. With this in view, a pebble bed test facility has been setup to study the heat transfer and pressure drop in pebble bed. Water is used as a working medium for the facility. The paper deals with the description of the pebble bed test facility and the experimental results of heat transfer and pressure drop. It also deals with the assessment of correlations for heat transfer and pressure drop in pebble bed geometry. Pressure drop experiments in the pebble bed test facility have been performed for Raynolds number ranges from 3000-12000. Various pressure drop correlations have been compared with the experimental data. It has been found that that the correlation given by Leva et. al. matches well with the experimental data. Various heat transfer correlations have also been compared. Heat transfer experiments are nearing completion

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

    International Nuclear Information System (INIS)

    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. Modularity of the MIT Pebble Bed Reactor for use by the commercial power industry

    OpenAIRE

    Hanlon-Hyssong, Jaime E.

    2008-01-01

    CIVINS The Modular Pebble Bed Reactor is a small high temperature helium cooled reactor that is being considered for both electric power and hydrogen production. Pebble bed reactors are being developed in South Africa, China and the US. To make smaller 120 Mwe reactors economically competitive with larger 1500 Mwe traditional light water reactors changes in the way these plants are built are needed. Economies of production need to be sufficiently large to compete with economies of sca...

  12. Neutronic simulation of a pebble bed reactor considering its double heterogeneous nature

    International Nuclear Information System (INIS)

    Highlights: ► A new model is successfully developed for a pebble bed reactor simulation. ► In the model, the double heterogeneous nature is considered using MCNP5 code. ► The initial and full core criticality, control rod worth, etc. are calculated to validate it. ► Results confirm the capability of Monte Carlo codes in modeling complex geometries. - Abstract: In pebble bed reactors, the core is filled with thousands of graphite and fuel pebbles. Fuel pebbles in these reactors consist of TRISO particles, which are embedded in a graphite matrix stochastically. The reactor core is also stochastically filled with pebbles. These two stochastic geometries comprise the so-called double heterogeneous nature of this type of reactor. In this paper, a pebble bed reactor, the HTR-10, is used to demonstrate a treatment of this double heterogeneity using the MCNP5 Monte Carlo code and MATLAB programming. In this technique, TRISO particles are modeled in a pebble using the expanded FILL and LATTICE features of MCNP5. MATLAB is used to generate random numbers which represent the location of pebbles in the core. Centers of pebbles are generated stochastically and uniformly and then transferred into the MCNP5 input file as the centers of spherical surfaces. In this model, there is no approximation to the actual geometry. In other words, the double heterogeneous nature is preserved while truncating neither the pebbles in the core nor the particles in the pebble matrix. Finally, to validate the model, benchmark problems of IAEA are used. Very good agreement with experimental results is observed.

  13. Effect of non-uniform porosity distribution on thermalhydraulics in a pebble bed reactor

    International Nuclear Information System (INIS)

    In pebble bed reactors, the porosity profile shows strong fluctuations near the wall. These changes in fuel density affect local power density, coolant velocity, and temperature distribution. This paper describes the pebFoam code, capable of calculating pebble bed thermohydraulics including non-uniform porosity distributions for arbitrary geometries, and investigates the changes in velocity, pressure drop, and helium and pebble temperatures when using a nonuniform porosity distribution instead of a uniform distribution. Results show only minor changes in temperature profiles and pressure drop for full power steady state calculations, though the velocity profile shows a clear increase in velocity near the wall. (author)

  14. Flow distribution of pebble bed high temperature gas cooled reactors using Large Eddy Simulation

    International Nuclear Information System (INIS)

    The simulation of complex three-dimensional gas flow through the gaps of the spherical fuel elements (fuel pebbles) of Pebble Bed Modulator Reactor is performed. This will help in understanding the highly three-dimensional, complex flow phenomena in pebble bed caused by flow curvature. The flow of this type has distinctive features, which strongly affect the boundary layer behavior. The transition from a laminar to turbulent flow around this curved flow occurs at different Reynolds (Re) numbers. Noncircular curved flows as in the pebble-bed situation need to be investigated. In this study, Large Eddy Simulation (LES) is used in modeling the turbulence to overcome the shortcoming of the Reynolds Average Navier-Stokes approach. (author)

  15. Optimization of OTTO Fuel Management in Pebble-Bed Reactors Using Particle Swarm Algorithm

    International Nuclear Information System (INIS)

    Pebble-Bed nuclear reactors feature highly flexible in-core fuel management capabilities due to on-line fueling and thermo-mechanical robust fuel design. Fuel pebbles with various fissile and fertile materials can be loaded into the reactor core at different rates. The fuel pebbles may be recirculated in the core several times until reaching their target burnup, or reach their target burnup in single pass through the core (OTTO- Once-Through-Then-Out fueling Scheme). Pebble-bed reactors have relatively efficient neutron economy since they operate with low excess reactivity and hence minimize the use of neutron poisons and control rods. Moreover, the fuel pebble robust design permits high burnup levels (up to 140000 MWD/THM). The flexibility of the fuel management operations allows enhancing fuel utilization. Traditionally fuel cycle design decisions were made using expert opinions and parametric studies. In this work, we have used the Particle Swarm Optimization (PSO) algorithm to optimize fuel utilization of pebble-bed reactors running OTTO fuel management. Optimization was carried out also for cores with Th232 as fertile material. Preliminary calculations were performed for a large core with 2 radial fuel loading zones. Results of the optimal fuel utilization performed for cores with UO2 fuel and cores with (Th- U)O2. Future work will include optimization of cores fuelled with separate seed (U) and blanket (Th) fuel pebbles and with advanced modular core configuration, like the PBMR400

  16. Characterisation of thermal radiation in the near-wall region of a packed pebble bed / Maritza de Beer

    OpenAIRE

    De Beer, Maritza

    2014-01-01

    The heat transfer phenomena in the near-wall region of a randomly packed pebble bed are important in the design of a Pebble Bed Reactor (PBR), especially when considering the safety case during accident conditions. At higher temperatures the contribution of the radiation heat transfer component to the overall heat transfer in a PBR increases significantly. The wall effect present in the near-wall region of a packed pebble bed affects the heat transfer in this region. Various correlations e...

  17. STUDI PEMODELAN DAN PERHITUNGAN TRANSPORT MONTE CARLO DALAM TERAS HTR PEBBLE BED

    Directory of Open Access Journals (Sweden)

    Zuhair .

    2013-01-01

    Full Text Available Konsep sistem energi VHTR baik yang berbahan bakar pebble (VHTR pebble bed maupun blok prismatik (VHTR prismatik menarik perhatian fisikawan reaktor nuklir. Salah satu kelebihan teknologi bahan bakar bola adalah menawarkan terobosan teknologi pengisian bahan bakar tanpa harus menghentikan produksi listrik. Selain itu, partikel bahan bakar pebble dengan kernel uranium oksida (UO2 atau uranium oksikarbida (UCO yang dibalut TRISO dan pelapisan silikon karbida (SiC dianggap sebagai opsi utama dengan pertimbangan performa tinggi pada burn-up bahan bakar dan temperatur tinggi. Makalah ini mendiskusikan pemodelan dan perhitungan transport Monte Carlo dalam teras HTR pebble bed. HTR pebble bed adalah reaktor berpendingin gas temperatur tinggi dan bermoderator grafit dengan kemampuan kogenerasi. Perhitungan dikerjakan dengan program MCNP5 pada temperatur 1200 K. Pustaka data nuklir energi kontinu ENDF/B-V dan ENDF/B-VI dimanfaatkan untuk melengkapi analisis. Hasil perhitungan secara keseluruhan menunjukkan konsistensi dengan nilai keff yang hampir sama untuk pustaka data nuklir yang digunakan. Pustaka ENDF/B-VI (66c selalu memproduksi keff lebih besar dibandingkan ENDF/B-V (50c maupun ENDF/B-VI (60c dengan bias kurang dari 0,25%. Kisi BCC memprediksi keff hampir selalu lebih kecil daripada kisi lainnya, khususnya FCC. Nilai keff kisi BCC lebih dekat dengan kisi FCC dengan bias kurang dari 0,19% sedangkan dengan kisi SH bias perhitungannya kurang dari 0,22%. Fraksi packing yang sedikit berbeda (BCC= 61%, SH= 60,459% tidak membuat bias perhitungan menjadi berbeda jauh. Estimasi keff ketiga model kisi menyimpulkan bahwa model BCC lebih bisa diadopsi dalam perhitungan HTR pebble bed dibandingkan model FCC dan SH. Verifikasi hasil estimasi ini perlu dilakukan dengan simulasi Monte Carlo atau bahkan program deterministik lainnya guna optimisasi perhitungan teras reaktor temperatur tinggi.   Kata-kunci: kernel, TRISO, bahan bakar pebble, HTR pebble bed

  18. Thermal-hydraulics numerical analyses of Pebble Bed Advanced High Temperature Reactor hot channel

    International Nuclear Information System (INIS)

    Background: The thermal hydraulics behavior of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) hot channel was studied. Purpose: We aim to analyze the thermal-hydraulics behavior of the PB-AHTR, such as pressure drop, temperature distribution of coolant and pebble bed as well as thermal removal capacity in the condition of loss of partial coolant. Methods: We used a modified FLUENT code which was coupled with a local non-equilibrium porous media model by introducing a User Defined Scalar (UDS) in the calculation domain of the reactor core and subjoining different resistance terms (Ergun and KTA) to calculate the temperature of coolant, solid phase of pebble bed and pebble center in the core. Results: Computational results showed that the resistance factor has great influence on pressure drop and velocity distribution, but less impact on the temperature of coolant, solid phase of pebble bed and pebble center. We also confirmed the heat removal capacity of the PB-AHTR in the condition of nominal and loss of partial coolant conditions. Conclusion: The numerical analyses results can provide a useful proposal to optimize the design of PB-AHTR. (authors)

  19. 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

  20. METHODS FOR MODELING THE PACKING OF FUEL ELEMENTS IN PEBBLE BED REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Abderrafi M. Ougouag; Joshua J. Cogliati; Jan-Leen Kloosterman

    2005-09-01

    Two methods for the modeling of the packing of pebbles in the pebble bed reactors are presented and compared. The first method is based on random generation of potential centers for the pebbles, followed by rejection of points that are not compatible with the geometric constraint of no (or limited) pebbles overlap. The second method models the actual physical packing process, accounting for the dynamic of pebbles as they are dropped onto the pebble bed and as they settle therein. A simplification in the latter model is the assumption of a starting point with very dilute packing followed by settling. The results from the two models are compared and the properties of the second model and the dependence of its results on many of the modeling parameters are presented. The first model (with no overlap allowed) has been implemented into a code to compute Dancoff factors. The second model will soon be implemented into that same code and will also be used to model flow of pebbles in a reactor and core densification in the simulation of earthquakes. Both methods reproduce experimental values well, with the latter displaying a high level of fidelity.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hans D. Gougar; Frederik Reitsma; Wessel Joubert

    2009-05-01

    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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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 233U inventory for a breeder could be produced by prebreeders using 235U fuel

  4. Absorber rod for nuclear reactors in a pebble bed of spherical operating elements

    International Nuclear Information System (INIS)

    The claim refers to the constructional configuration of an absorber rod, whose and penetrating into the pebble bed has an opening to reduce the fracture rate, so that the operating elements can escape into a channel within the absorber rod. To suit this to the direction of movement of the elements a part of the end of the rod is flexibly connected to the hollow absorber rod via a joint. In this way the mechanical load of the element particles is reduced and simultaneously one achieves that much lower force is required to insert the absorber rod into the pebble bed. (UA)

  5. Prototype studies on the nondestructive online burnup determination for the modular pebble bed reactors

    International Nuclear Information System (INIS)

    Highlights: • Prototype study of online burnup measurement for HTR proves its feasibility. • Calibration and its correction of burnup assay device is discussed and verified. • Analysis of simulated gamma spectra shows good performance of spectra-unfolding method. - Abstract: The online fuel pebble burnup determination in future modular pebble bed reactor is implemented by measuring nondestructively the activity of the monitoring nuclide Cs-137 with HPGe detector on a pebble-by-pebble basis. Based on a full size prototype the feasibility is investigated. The prototype was first tested by using double sources to show that a precision of 2.8% (1σ) can be achieved in the determination of the Cs-137 net counting rate. Then, the relationship between the Cs-137 activity and the net counting rate recorded in the HPGe detector is calibrated with a standard Cs-137 source contained in the center of a graphite sphere with the same dimension as a real fuel pebble. Because the self attenuation of the calibration source differs with a fuel pebble, a correction factor of 1.07 ± 0.02 (p = 0.95) to the calibration is derived by using the efficiency transfer method. Last, by analyzing the spectra generated with KORIGEN software followed by Monte Carlo simulation, it is predicted that the relative standard deviation of the Cs-137 net counting rate can be still controlled below 3.5% despite of the presence of all the interfering peaks. The results demonstrate the feasibility of utilizing HPGe gamma spectrometry in the online determination of the pebble burnup in future modular pebble bed reactors

  6. Measurement of the thermal conductivity and heat transfer coefficient of a binary bed of beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Donne, M.D.; Piazza, G. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Goraieb, A.; Sordon, G.

    1998-01-01

    The four ITER partners propose to use binary beryllium pebble bed as neutron multiplier. Recently this solution has been adopted for the ITER blanket as well. In order to study the heat transfer in the blanket the effective thermal conductivity and the wall heat transfer coefficient of the bed have to be known. Therefore at Forschungszentrum Karlsruhe heat transfer experiments have been performed with a binary bed of beryllium pebbles and the results have been correlated expressing thermal conductivity and wall heat transfer coefficients as a function of temperature in the bed and of the difference between the thermal expansion of the bed and of that of the confinement walls. The comparison of the obtained correlations with the data available from the literature show a quite good agreement. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Pohl, P. [Arbeitsgemeinschaft Versuchsreaktor AVR GmbH, Postfach 1160, 52412 Juelich (Germany)

    2006-07-01

    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)

  8. Temperature transients of a fusion-fission ITER pebble bed reactor in loss of coolant accident

    International Nuclear Information System (INIS)

    In this preliminary scoping study, post-accident temperature transients of several fusion-fission designs utilizing ITER-FEAT-like parameters and fission pebble bed fuel technology are examined using a 1-D cylindrical MATLAB heat transfer code along with conventional fission decay heat approximations. Scenarios studied include systems with no additional passive safety features to systems with melting reflectors designed to increase emissivity after reaching a specified temperature. Results show that for a total fission power of ∼1400-2800 MW, two of the realistic variants investigated are passively safe. The crucial time, defined as the time when either any structural part of the fusion-fission tokamak reaches melting point, or when the pebble fuel reaches 1873 K, ranges from 5.7 to 76 h for the unsafe configurations. Additionally, it is illustrated that, fundamentally, the LOCA characteristics of pure fission pebble beds and fusion-fission pebble beds are different. Namely, the former depends on the pebble fuel's large thermal capacity, along with external radiation and natural convective cooling, while the latter depends significantly more on the tokamak's sizeable total internal heat capacity. This difference originates from the fusion-fission reactor's conflicting goal of having to minimize heat transfer to the magnets during normal operation. These results are discussed in the context of overall fusion-fission reactor design and safety

  9. 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

  10. Proliferation Resistant Fuel for Pebble Bed Modular Reactors

    International Nuclear Information System (INIS)

    Proliferation of nuclear weapons produced with power reactors plutonium has always been amajor problem of the nuclear energy industry. This includes the PebbleBed Modular Reactor(PBMR), which is a specific design of a GenIV High-Temperature Reactor (HTR), mainly due to its online refueling feature, which may be misused for the production of weapons gradeplutonium. A promising approach toward preventing the proliferation of power reactorplutoniumis to denaturate the plutonium by increasing the ratio of 238Pu to total Pu in the spentfuel(1). The 238Pu isotope is characterized by a high heat rate (approximately 567 W/kg) due to thealpha decay of the 238Pu with half-life of 87.74 yr, in addition to its high spontaneous fissionneutron emission, which is higher than that of 240Pu. Thus, the presence of 238Pu in Pu considerably complicates the design and construction of nuclear weapons based on Pu, owing tothese characteristics of 238Pu. Recent papers(2,3) show that a Pu mixture is proliferation resistant given that the weight ratio of 238Pu to Pu is larger than 6%. In this paper we have studied afeasible technique for ensuring that the 238Pu to Pu ratio, in the Pu produced in PBMR, is larger than 6% during the entire fuel cycle. Contamination of the spent fuel with 238Pu may be achieved by doping the nuclear fuel witheither 241Am or 237Np(4-13). The 238Pu isotope is obtained from both 241Am and 237Np by a neutron-capture reaction and the subsequent decay of the reaction products(13).The 237Np isotopeis by itself a potential weapons grade material. However, its large critical mass of 57±4 kg(14) andthe difficulty of extracting it from irradiated fuel elements make it impractical for weapons purposes. On the other hand, the critical mass of 241Am is smaller, i.e. 34 to 45 kg. However, withdecay heat production of 114W/kg, the critical mass becomes a heat source of 3.9 to 5.1 KW,which makes 241Am unsuitable for weapons applications(3). As a result, it is a non

  11. CORE ANALYSIS, DESIGN AND OPTIMIZATION OF A DEEP-BURN PEBBLE BED REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-05-01

    Achieving a high burnup in the Deep-Burn pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. To investigate the aforementioned effects a code system using existing codes has been developed for neutronic, thermal-hydraulic and fuel depletion analysis of Deep-Burn pebble bed reactors. A core analysis of a Deep-Burn Pebble Bed Modular Reactor (400 MWth) design has been performed for two Deep-Burn fuel types and possible improvements of the design with regard to power peaking and temperature reactivity feedback are identified.

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

    International Nuclear Information System (INIS)

    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)

  13. 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

  14. In-core fuel management optimization of a Very High Temperature pebble-bed Reactor

    International Nuclear Information System (INIS)

    A new calculation procedure was developed to reduce the power peak in the core of a Very High Temperature pebble-bed Reactor. The procedure consists in several coupled computational codes, which are used iteratively until convergence is reached. This procedure combines the fuel depletion and the neutronic behavior of the fuel in the reactor core, modeling once-through-then-out cycles as well as cycles in which pebbles are recirculated through the core an arbitrary number of times, obtaining the asymptotic fuel-loading pattern directly, without any intermediate loading pattern. (Author)

  15. 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.

  16. Quasi-direct numerical simulation of a pebble bed configuration, Part-II: Temperature field analysis

    International Nuclear Information System (INIS)

    Highlights: ► Quasi direct numerical simulations (q-DNSs) of a pebble bed configuration have 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 of temperature and respective turbulent heat fluxes are extensively reported in this paper. -- Abstract: Good prediction of the flow and heat transfer phenomena in the pebble bed core of a high temperature reactor (HTR) is a challenge for available turbulence models, which still require to be validated. While experimental data are generally desirable in this validation process, due to the complex geometric configuration and measurement difficulties, a very limited amount of data is currently available. On the other hand, direct numerical simulation (DNS) is considered an accurate simulation technique, which may serve as an alternative for validating turbulence models. In the framework of the present study, quasi-direct numerical simulation (q-DNS) of a single face cubic centered pebble bed is performed, which will serve as a reference for the validation of different turbulence modeling approaches in order to perform calculations for a randomly arranged pebble bed. These simulations were performed at a Reynolds number of 3088, based on pebble diameter, with a porosity level of 0.42. Results related to flow field (mean, RMS and covariance of velocity) have been presented in Part-I, whereas, in the present article, we focus our attention to the analysis of the temperature field. A wide range of qualitative and quantitative data for the thermal field (mean, RMS and turbulent heat flux) has been generated

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

    International Nuclear Information System (INIS)

    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

  18. Numerical simulation on friction coefficient effect of pebble flow dynamics in two-dimensional pebble-bed reactor

    International Nuclear Information System (INIS)

    In order to investigate the pebble flow dynamics in the high-temperature reactor core and based on the two-dimensional experiments of pebble flow dynamics, discrete element method (DEM) was used to simulate the pebble flow dynamics. The mean flow stream lines, standard deviation and the mean residence time of the pebble flow zone generated by markers were compared and analyzed. The results show that ball friction coefficient has little effect on the pebble flow field. With the pebble friction coefficient increasing, the horizontal diffusion of pebbles decreases and the pebble flow seems to be more uniform. The wall friction coefficient has little effect on the horizontal diffusion. While the wall friction coefficient increases, the flow tends to be more uneven. (authors)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  1. An analysis of the thermal behaviour of pebble-bed nuclear reactors in the case of emergencies

    International Nuclear Information System (INIS)

    In this paper, the performance of pebble-bed nuclear reactors under very severe emergencies will be analysed. Calculated hypotheses take into consideration total failure of decay heat removal systems and any other active equipment, including electric power supply. It has been shown that pebble temperatures will remain well below safety limits if the reactor design embodies a core catcher with a passive cooling reservoir and a pebble draining system which would be naturally activated by a lack of a power supply. Although these features apply to any pebble-bed reactor, particular attention is paid to accelerator-driven sub-critical assemblies, where reactivity noise produced pebble quivering has a practical negligible effect. (authors)

  2. Gas Reactor International Cooperative Program: German Pebble Bed Reactor Technology review update

    International Nuclear Information System (INIS)

    This report provides a review of the German pebble bed reactor technology, and updates the information provided in the Gas Reactor International Cooperative Program Interim Report COO-4057-6, German Pebble Bed Reactor Design and Technology Review, dated September 1978. Most of the updated information is for the PNP-500 and the HHT-Prototype plants. The PNP-500 is a 500 MW(t) multi-purpose demonstration plant for coal conversion applications. The HHT-Prototype is a 1640 MWt reactor designed to produce 675 MWe of electricity using a direct cycle gas turbine. The report provides a description and evaluation of the overall plant and the nuclear reactor for both the PNP-500 and HHT-Prototype. A description and evaluation of the primary system components is presented for the process heat and gas turbine applications

  3. Transuranics elimination in an optimised pebble-bed sub-critical reactor

    International Nuclear Information System (INIS)

    In a nuclear energy economy the nuclear waste is a big burden to its further development and deployment. The possibility of eliminating the long-term part of the waste presents an appealing opportunity to the sustainability and acceptance of a better and cleaner source of energy. It is shown that the proposed pebble-bed transmutator has suitable characteristics to transmute most of the isotopes that contribute to the long-term radioactivity. This proposed reactor presents also inherent safety characteristics, which is a necessary element in a new reactor design to be accepted by the society. Throughout this paper, we will characterise the new reactor concept, and present some of the neutronics and safety characteristics of an accelerator driven pebble-bed reactor, (ADS) for transuranics elimination. (author)

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

    International Nuclear Information System (INIS)

    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

  5. Automated control of a pebble bed core thermal flow test unit / by Jan H.J. Prinsloo

    OpenAIRE

    Prinsloo, Jan Hendrik Jacobus

    2006-01-01

    The HTTF (Heat Transfer Test Facility) is a unique project verifying the only pebble bed correlations currently used by PBMR (Pty) LTD. They are developing a new concept nuclear power station and are at present in the preparation phase of the conshuction of the worlds first PBMR (Pebble Bed Modular Reactor). The PBMR required the HTTF to be built at the North-West University in Potchefstroom. The HTTF consists of two separate test facilities: the H7TU (High Temperature Test Uni...

  6. Modular Pebble-Bed Reactor Project: Laboratory-Directed Research and Development Program FY 2002 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew; Dolan, Thomas James; Miller, Gregory Kent; Moore, Richard Leroy; Terry, William Knox; Ougouag, Abderrafi Mohammed-El-Ami; Oh, Chang H; Gougar, Hans D

    2002-11-01

    This report documents the results of our research in FY-02 on pebble-bed reactor technology under our Laboratory Directed Research and Development (LDRD) project entitled the Modular Pebble-Bed Reactor. The MPBR is an advanced reactor concept that can meet the energy and environmental needs of future generations under DOE’s Generation IV initiative. Our work is focused in three areas: neutronics, core design and fuel cycle; reactor safety and thermal hydraulics; and fuel performance.

  7. Numerical Simulation of Accident Scenario in High Temperature Gas Cooled (Pebble Bed) Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Peter, Geoffrey J. [Oregon Institute of Technology - Portland Center, Portland (United States)

    2012-03-15

    The accident scenario resulting from blockages due to the retention of dust in the coolant gas or from the rupture of one or more fuel particles used in the High Temperature Gas Cooled (Pebble Bed) Nuclear Reactors is considered in this paper. The next generation of Advanced High Temperature Reactors (AHTR), are considered for nuclear power production, and for high-temperature hydrogen production using nuclear reactors to reduce the carbon footprint. Blockages can cause LOCA variations in flow and heat transfer that may lead to hot spots within the bed that could compromise reactor safety. Therefore, it is important to know the void fraction distribution and the interstitial velocity field in the packed bed. The blockage for this numerical study simulated a region with significantly lower void than that in the rest of the bed. Finite difference technique solved the simplified continuity, momentum, and energy equations. Any meaningful outcome of the solution depended largely upon the validity of the boundary conditions. Among them, the inlet and outlet velocity profiles required special attention. Thus, a close approximation to these profiles obtained from an experimental set-up established the boundary conditions. This paper presents the development of the elliptic-partial equation for a bed of a bed of pebbles, and the solution procedure. The paper also discusses velocity and temperature profiles obtained from both numerical and experimental set-up, with and without effect of blockage. Based on the studies it is evident that knowledge of LOCA velocity and temperature distribution within the fuel element in a Pebble Bed Nuclear Reactor or AHTR is essential for reactor safety.

  8. Medium voltage direct current (MVDC) converter for pebble bed modular reactor (PBMR) / Hendrik de Villiers Pretorius

    OpenAIRE

    Pretorius, Hendrik de Villiers

    2004-01-01

    Nuclear and renewable energy systems will probably be used more and more extensively in future due to high environmental demands regarding pollution and exhaustion of the world's gas and coal reserves. Because most types of renewable energy systems do not supply electric power at line frequency and voltage a converter is used to connect these sources to the existing power system. The Pebble Bed Modular Reactor (PBMR) is a nuclear power plant currently using a 50 Hz synchrono...

  9. Design of a power conversion system for an indirect cycle, helium cooled pebble bed reactor system

    International Nuclear Information System (INIS)

    A design is presented for the turbomachinery for an indirect cycle, closed, helium cooled modular pebble bed reactor system. The design makes use of current technology and will operate with an overall efficiency of 45%. The design uses an intermediate heat exchanger which isolated the reactor cycle from the turbomachinery. This design excludes radioactive fission products from the turbomachinery. This minimizes the probability of an air ingress accident and greatly simplifies maintenance. (author)

  10. Generic Investigations on Transport Theory Modelling of High Temperature Reactors of Pebble Bed Type

    OpenAIRE

    Sureda Sureda, Antonio Jaime

    2008-01-01

    The GRS (Gesellschaft fuer Anlagen und Reaktorsicherheit = Company for Plant and Reactor Safety) maintains and further develops the code system DORT-TD/HERMIX-DIREKT, which is a complex tool for the simulation of coupled neutronics/thermal-hydraulics transients and accident scenarios of high-temperature gas cooled reactors of pebble bed type. With this tool, GRS takes part in the international benchmark activity "OECD/NEA PBMR400 Transient Benchmark”, which aims at the simulation of transient...

  11. The study on fuel effect in discharge pipe of pebble bed reactor

    International Nuclear Information System (INIS)

    The simulation method of fuel loading in discharge pipe of pebble bed reactor is introduced. As an exemplary application case, the effect of fuel elements in the discharge pipe on reactor physics and thermohydraulic properties is calculated and analysed by CHTRP code in HTR-10 MW. The calculation gives the power and temperature distribution in the area of the discharge pipe, very useful for further analysis of reactor physics and safety

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

    International Nuclear Information System (INIS)

    Highlights: • A 300 MWt 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 UO2 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 UO2 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 UO2 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

  13. Proposal for an international experimental pebble bed reactor - HTR2008-58174

    International Nuclear Information System (INIS)

    HTRs, both prismatic block fuelled and pebble fuelled, feature a number of uniquely beneficial characteristics that will be discussed in this paper. In this paper the construction of an international experimental pebble bed reactor is proposed, possible experiments suggested and an invitation extended to interested partners for co-operation in the project. Experimental verification by nuclear regulators in order to facilitate licensing and the development of a new generation of reactors create a strong need for such a reactor. Suggested experiments include: Optimized incineration of waste Pu in a pebble bed reactor: The capability to incineration pure reactor grade plutonium by means of ultra high burn-up in pebble bed reactors will be presented at this conference in the track on fuel and fuel cycles. This will enable incineration of the global stockpile of separated reactor grade Pu within a relatively short time span. Testing of fuel sphere geometries, aimed at improving neutron moderation and a decrease in fuel temperatures. Th/Pu fuel cycles: Previous HTR programs demonstrated the viability of a Th-232 fuel-cycle, using highly enriched uranium (HEU) as driver material. However, considerations favoring proliferation resistance limit the enrichment level of uranium in commercial reactors to 20 %, thereby lowering the isotopic efficiency. Therefore, Pu driver material should be developed to replace the HEU component. Instead of deploying a (Th, Pu)O fuel concept, the proposal is to use the unique capability offered by pebble bed reactors in deploying separate Th- and Pu-containing pebbles, which can be cycled differently. Testing of carbon-fiber-carbon (CFC) structures for in-core or near-core applications, such as guide tubes for reserve shutdown systems, thus creating the possibility to safely shutdown reactors with increased diameter. Development of very high temperature reactor components for process heat applications. Advanced decay heat removal systems e

  14. Development and testing of analytical models for the pebble bed type HTRs

    International Nuclear Information System (INIS)

    The pebble bed type gas cooled high temperature reactor (HTR) appears to be a good candidate for the next generation nuclear reactor technology. These reactors have unique characteristics in terms of the randomness in geometry, and require special techniques to analyze their systems. This study includes activities concerning the testing of computational tools and the qualification of models. Indeed, it is essential that the validated analytical tools be available to the research community. From this viewpoint codes like MCNP, ORIGEN and RELAP5, which have been used in nuclear industry for many years, are selected to identify and develop new capabilities needed to support HTR analysis. The geometrical model of the full reactor is obtained by using lattice and universe facilities provided by MCNP. The coupled MCNP-ORIGEN code is used to estimate the burnup and the refuelling scheme. Results obtained from Monte Carlo analysis are interfaced with RELAP5 to analyze the thermal hydraulics and safety characteristics of the reactor. New models and methodologies are developed for several past and present experimental and prototypical facilities that were based on HTR pebble bed concepts. The calculated results are compared with available experimental data and theoretical evaluations showing very good agreement. The ultimate goal of the validation of the computer codes for pebble bed HTR applications is to acquire and reinforce the capability of these general purpose computer codes for performing HTR core design and optimization studies

  15. Coupling of RMC and CFX for analysis of Pebble Bed-Advanced High Temperature Reactor core

    International Nuclear Information System (INIS)

    Highlights: ► The CFD code CFX is used for whole pebble bed reactor core calculation. ► The Monte Carlo Code RMC and CFX are used for the coupling of neutronics and T-H. ► Coupled calculations for steady-state problem can reach stable results. ► Increasing the number of neutron histories is effective to improve accuracy. - Abstract: This paper introduces a steady-state coupled calculation method using the Monte Carlo Code RMC (Reactor Monte Carlo) and the Computational Fluid Dynamic (CFD) code CFX for the analysis of a Pebble Bed-Advanced High Temperature Reactor (PB-AHTR) core. The RMC code is used for neutronics calculation while CFX is used for Thermal-Hydraulics (T-H) calculation. The porous media model is used in CFX modeling to simulate the pebble bed structure in PB-AHTR. The CFX model has also been validated against the RELAP5-3D model developed in the previous research. The script language PERL is used as a development tool to manipulate and control the entire coupled calculation. This research gives the conclusion that the steady-state coupled calculation using RMC and CFX is feasible and can obtain stable results within a few iterations. However, due to the statistical errors of Monte Carlo method, the fluctuation of results still occurs. For the purpose of improving the accuracy, the paper applies and discusses two methods, of which increasing the number of neutron histories is an effective method.

  16. 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.

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

    International Nuclear Information System (INIS)

    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.

  20. Computational and experimental prediction of dust production in pebble bed reactors—Part I

    International Nuclear Information System (INIS)

    Highlights: • A nonlinear dimensionless wear coefficient is theoretically proposed. • A material constant for the relation of asperity height and wear is introduced. • A nonlinear modification of Archard wear formula is proposed. • The graphite wear dust production in a typical pebble bed reactor is predicted. • Experimental and computational wear results for graphite are presented. -- Abstract: This paper describes the computational modeling and simulation, and experimental testing of graphite moderators in frictional contacts as anticipated in a pebble bed reactor. The potential of carbonaceous particulate generation due to frictional contact at the surface of pebbles and the ensuing entrainment and transport into the gas coolant are safety concerns at elevated temperatures under accident scenarios such as air ingress in the high temperature gas-cooled reactor. The safety concerns are due to the documented ability of carbonaceous particulates to adsorb fission products and transport them in the primary circuit of the pebble bed reactor, thus potentially giving rise to a relevant source term under accident scenarios. Here, a finite element approach is implemented to develop a nonlinear wear model in air environment. In this model, material wear coefficient is related to the changes in asperity height during wear. The present work reports a comparison between the finite element simulations and the experimental results obtained using a custom-designed tribometer. The experimental and computational results are used to estimate the quantity of nuclear grade graphite dust produced from a typical anticipated configuration. In Part II, results from a helium environment at higher temperatures and pressures are experimentally studied

  1. Monte Carlo Calculations of Pebble Bed Benchmark Configurations of the PROTEUS Facility

    International Nuclear Information System (INIS)

    Under the auspices of the International Atomic Energy Agency, a series of well-documented benchmark experiments were performed at the Proteus facility of the Swiss Paul Scherrer Institute. Thirteen critical pebble bed reactor configurations were assembled, with ten of them deterministic with a precise location of the low-enriched fuel and moderator pebbles. Seven of these configurations were modeled with a very high spatial resolution with the Monte Carlo code MCNP with details that go from the fuel kernel (0.5 mm in diameter) to the walls surrounding the facility. The calculations of the k's of the configurations agree quite well with the experiments (within a fraction of a dollar). A sensitivity analysis is included to discuss the possibility of a small bias; also biases introduced by customary approximations of production codes were calculated. The experiments and the analysis of this paper might be very useful tools to check the calculational accuracy of procedures used in the emerging work related to pebble bed modular gas-cooled reactors

  2. 3D DEM simulation and analysis of void fraction distribution in a pebble bed high temperature reactor

    International Nuclear Information System (INIS)

    Highlights: • We show a detailed analysis of void fraction (VF) in HTR-10 of China using DEM. • Radial distribution (RD) of VF is uniform in the core and oscillated near the wall. • Axial distribution (AD) is linearly varied along height due to effect of gravity. • Steady RD of VF in the conical base is Gaussian-like, larger than packing bed. • Joint linear and normal distribution of VF is analyzed and explained. - Abstract: The current work analyzes the radial and axial distributions of void fraction of a pebble bed high temperature reactor. A three-dimensional pebble bed corresponding to our test facility of pebble bed type gas-cooled high temperature reactor (HTR-10) in Tsinghua University is simulated via discrete element method, and the radial and axial void fraction profiles are calculated. It validates the oscillating characteristics of radial void fraction near the wall. Detailed calculations show the differences of void fraction profiles between the stationary packing bed and the dynamically discharging bed. Based on the vertically and circumferentially averaged radial distribution and horizontally averaged axial distribution of void fraction, a fully three-dimensional analytical distribution of void fraction throughout the bed is established. The results show the combined effects of gravity and void variation in the pebble bed caused by the pebble discharging. It indicates the linearly increased packing effect caused by gravity in the vertical (axial) direction and the normal distribution of void in the horizontal (radial) direction by pebble drainage. These two effects coexist in the conical base of the bed whereas only the former effect exists in the cylindrical volume of the bed

  3. Analysis of dust and fission products in a pebble bed NGNP

    International Nuclear Information System (INIS)

    In the HTR pebble bed reactor graphite dust is generated during normal reactor operation due to pebble-to-pebble interactions. This dust will be deposited throughout the primary system. Furthermore, the dust will become radioactive due to sorption of fission products released during normal operation. This paper presents an analysis of dust transport and deposition and fission product transport and plate-out during normal operation of a NGNP (next generation nuclear plant) Westinghouse pebble bed design. The main objective is to determine the amount and location of deposited graphite dust in the system and the amount of radioactive isotopes adsorbed on the structures and the dust during normal operation. The results will be used in planning of maintenance activities. Moreover the present results may be used in a next step to perform a depressurized loss of forced cooling (D-LOFC) analysis and to determine the amount of radioactivity released to the atmosphere during the accident. The analysis was performed using the SPECTRA code. Based on the assumed dust source term, during the 60 years lifetime of the pebble bed reactor concept which was analyzed approximately 1630 kg dust enters the primary helium flow. It was determined: (1)That 86% of the dust settles on the graphite structures inside the reactor vessel. Of the remaining graphite dust 2/3 collects in the low temperature intermediate heat exchanger (IHX) and the remaining dust in the high temperature IHX and the connecting pipes of the primary system. (2)Agglomeration of dust from smaller into larger particles is observed at all locations. The resuspension of agglomerated dust particles is an important phenomenon that limits the build-up of a dust layer on the surface of the IHXs. A large deposited dust layer is observed on the control rod drive (CRD) walls in the graphite structures. The dust accumulates on the CRD walls because, due to very low gas velocities, resuspension of the agglomerated particles does

  4. Optimal study of a solar air heating system with pebble bed energy storage

    International Nuclear Information System (INIS)

    Highlights: → Use two kinds of circulation media in the solar collector. → Air heating and pebble bed heat storage are applied with different operating modes. → Design parameters of the system are optimized by simulation program. → It is found that the system can meet 32.8% of the thermal energy demand in heating season. → Annual solar fraction aims to be 53.04%. -- Abstract: The application of solar air collectors for space heating has attracted extensive attention due to its unique advantages. In this study, a solar air heating system was modeled through TRNSYS for a 3319 m2 building area. This air heating system, which has the potential to be applied for space heating in the heating season (from November to March) and hot water supply all year around in North China, uses pebble bed and water storage tank as heat storage. Five different working modes were designed based on different working conditions: (1) heat storage mode, (2) heating by solar collector, (3) heating by storage bed, (4) heating at night and (5) heating by an auxiliary source. These modes can be operated through the on/off control of fan and auxiliary heater, and through the operation of air dampers manually. The design, optimization and modification of this system are described in this paper. The solar fraction of the system was used as the optimization parameter. Design parameters of the system were optimized by using the TRNSYS program, which include the solar collector area, installation angle of solar collector, mass flow rate through the system, volume of pebble bed, heat transfer coefficient of the insulation layer of the pebble bed and water storage tank, height and volume of the water storage tank. The TRNSYS model has been verified by data from the literature. Results showed that the designed solar system can meet 32.8% of the thermal energy demand in the heating season and 84.6% of the energy consumption in non-heating season, with a yearly average solar fraction of 53.04%.

  5. Transmutation of nuclear wastes with gas-cooled pebble-bed ads

    International Nuclear Information System (INIS)

    Transmutation of nuclear wastes is being explored for its application to waste management, a fundamental issue for nuclear industry. Several concepts are under consideration, mainly fast breeder reactors and accelerator driven systems (ADS). Inside this second category, we are analysing a helium-cooled graphite moderated sub-critical assembly, which uses as fuel units a small amount of transuranics diluted, in the form of TRISO coated particles, in graphite pebbles. This configuration (PBT) allows for neutron spectra that, taking advantage of the existence of huge capture resonances in the epithermal region, increase in a substantial factor the system transmutation efficiency. Neutronic studies to determine transmutation performance and thermal behaviour are presented and discussed together with an analysis of the additional studies to address before going into detailed design activities. (author)

  6. The preliminary analysis on the steady-state and kinetic features of the molten salt pebble-bed reactor

    International Nuclear Information System (INIS)

    A novel design concept of molten salt pebble-bed reactor with an ultra-simplified integral primary circuit called 'Nuclear Hot Spring' has been proposed, featured by horizontal coolant flow in a deep pool pebble-bed reactor, providing 'natural safety' features with natural circulation under full power operation and less expensive primary circuit arrangement. In this work, the steady-state physical properties of the equilibrium state of the molten salt pebble-bed reactor are calculated by using the VSOP code, and the steady-state thermo-hydraulic analysis is carried out based on the approximation of absolutely horizontal flow of the coolant through the core. A new concept of 2-dimensional, both axial and radial, multi-pass on-line fuelling scheme is presented. The result reveals that the radial multi-pass scheme provides more flattened power distribution and safer temperature distribution than the one-pass scheme. A parametric analysis is made corresponding to different pebble diameters, the key parameter of the core resistance and the temperature at the pebble center. It is verified that within a wide range of pebble diameters, the maximum pebble center temperatures are far below the safety limit of the fuel, and the core resistance is considerably less than the buoyant force, indicating that the natural circulation under full power operation is achievable and the ultra-simplified integral primary circuit without any pump is possible. For the kinetic properties, it is verified that the negative temperature coefficient is achieved in sufficient under-moderated condition through the preliminary analysis on the temperature coefficients of fuel, coolant and moderator. The requirement of reactivity compensation at the shutdown stages of the operation period is calculated for the further studies on the reactivity control. The molten salt pebble-bed reactor with horizontal coolant flow can provide enhanced safety and economical features. (authors)

  7. Potential and limitations in maximizing the power output of an inherent safe modular pebble bed HTGR

    International Nuclear Information System (INIS)

    The past development of modular pebble-bed HTGRs in Germany led to two well-defined reactor designs, namely the 200 MWth HTR-MODUL and the 250 MWth HTR-100 by SIEMENS/Interatom and ABB/HRB, respectively. Recently the South African utility, ESKOM, decided to include the pebble-bed HTGR design as a future supply option. In contrast to the German designs, ESKOM prefers a direct cycle helium turbine system on the power conversion side. This imposes certain modified boundary conditions on the reactor design and enables a higher plant efficiency. Nuclear and thermal-hydraulic investigations have been performed at KFA-ISR to determine the potential and limitations of increasing the unit thermal power output of the reactor compared to the former german designs. In doing so an upper limit for the maximum fuel element temperature of 1600 deg. C was observed. The impact of all modifications in view onto the efficiency of the nuclear control and shut-down systems was also considered. The results obtained so far demonstrate the well-adapted and conservative design of the SIEMENS HTR-MODUL within a 10% safety margin to the higher region. The introduction of graphite noses has a remarkably positive influence on the shut-down and control systems, while the positive effect on the maximum accident temperature depends strongly on the fast neutron dose-related thermal conductivity of the nose graphite. Considering the fact that effective conductivity of the pebble-bed core is maintained at high temperatures, the temperature effect due to the noses are of secondary influence at this point. (author)

  8. 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

  9. 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

    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

  10. Development of a 3D multigroup program for Dancoff factor calculation in pebble bed reactors

    International Nuclear Information System (INIS)

    Highlights: • Development of a 3D Monte Carlo based code for pebble bed reactors. • Dancoff sensitivity to clad, moderator and fuel cross sections is considered. • Sensitivity of Dancoff to number of energy groups is considered. • Sensitivity of Dancoff to number of fuel and their arrangement is considered. • Excellent agreements vs. MCNP code. - Abstract: The evaluation of multigroup constants in reactor calculations depends on several parameters. One of these parameters is the Dancoff factor which is used for calculating the resonance integral and flux depression in the resonance region in heterogeneous systems. In the current paper, a computer program (MCDAN-3D) is developed for calculating three dimensional black and gray Dancoff coefficients, based on Monte Carlo, escape probability and neutron free flight methods. The developed program is capable to calculate the Dancoff factor for an arbitrary arrangement of fuel and moderator pebbles. Moreover this program can simulate fuels with homogeneous and heterogeneous compositions. It might generate the position of Triso particles in fuel pebbles randomly as well. It could calculate the black and gray Dancoff coefficients since fuel region might have different cross sections. Finally, the effects of clad and moderator are considered and the sensitivity of Dancoff factor with fuels arrangement variation, number of TRISO particles and neutron energy has been studied

  11. Safety assessment of the helium-cooled pebble bed test blanket module for ITER

    International Nuclear Information System (INIS)

    The European helium-cooled pebble bed blanket is one of six candidates to be tested in ITER. The corresponding test module and cooling system have been analysed for off-normal accident scenarios, involving large in-vessel and ex-vessel coolant leaks, leaks inside the module, and complete loss of flow. The methods involve transient systems analyses, local FE temperature analyses, 1 D heat transport calculations and chemical reaction estimates. Results are summarised with view to pressure evolution in ITER compartments, short and long-term temperature history in the module, decay heat removal and chemical reaction rates. (authors)

  12. Effects of Spatial Variations in Packing Fraction on Reactor Physics Parameters in Pebble-Bed Reactors

    International Nuclear Information System (INIS)

    The well-known spatial variation of packing fraction near the outer boundary of a pebble-bed reactor core is cited. The ramifications of this variation are explored with the MCNP computer code. It is found that the variation has negligible effects on the global reactor physics parameters extracted from the MCNP calculations for use in analysis by diffusion-theory codes, but for local reaction rates the effects of the variation are naturally important. Included is some preliminary work in using first-order perturbation theory for estimating the effect of the spatial variation of packing fraction on the core eigenvalue and the fision density distribution

  13. Effects of Spatial Variations in Packing Fraction of Reactor Physics Parameters in Pebble-Bed Reactors

    International Nuclear Information System (INIS)

    The well-known spatial variation of packing fraction near the outer boundary of a pebble-bed reactor core is cited. The ramifications of this variation are explored with the MCNP computer code. It is found that the variation has negligible effects on the global reactor physics parameters extracted from the MCNP calculations for use in analysis by diffusion-theory codes, but for local reaction rates the effects of the variation are naturally important. Included is some preliminary work in using first-order perturbation theory for estimating the effect of the spatial variation of packing fraction on the core eigenvalue and the fission density distribution

  14. Numerical analysis of dynamic behavior of HTR pebble-bed core and comparison with test results

    International Nuclear Information System (INIS)

    The behavior under seismic loading of the pebble bed core of a high temperature reactor is the objective of the investigation reported here. The paper describes the constitutive modelling of the assembly of spheres comprising the core and the finite element simulation of shaking table tests conducted on a one-sixth physical model of the core of a proposed new medium-sized HTR power plant. The analytical studies and the shaking table tests have been performed with the aim of gaining a fundamental understanding of the dynamic behavior of such core material and validating numerical models

  15. The pebble bed high temperature reactor as a source of nuclear process heat. Vol. 2

    International Nuclear Information System (INIS)

    A theoretical analysis is given for a series of 8 different variants of the pebble-bed reactor in the 'once through' fuel management scheme. The comparison gives some insight into the parametric sensitivities and into the development potential of this type. The thorium/U-233 recycling fuel cycle allows to increase the conversion ratio up to the range between 0.90 and 0.95. The feasibility for a changeover between different fuel cycles under full power operation. - The study is complemented by a review of the relevant previous investigations. (orig.)

  16. Effects of Spatial Variations in Packing Fraction on Reactor Physics Parameters in Pebble-Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    William K. Terry; A. M. Ougouag; Farzad Rahnema; Michael Scott McKinley

    2003-04-01

    The well-known spatial variation of packing fraction near the outer boundary of a pebble-bed reactor core is cited. The ramifications of this variation are explored with the MCNP computer code. It is found that the variation has negligible effects on the global reactor physics parameters extracted from the MCNP calculations for use in analysis by diffusion-theory codes, but for local reaction rates the effects of the variation are naturally important. Included is some preliminary work in using first-order perturbation theory for estimating the effect of the spatial variation of packing fraction on the core eigenvalue and the fision density distribution.

  17. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into 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 (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 239-Pu, 240-Pu and 241-Pu 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. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge

  18. Conceptual design study of Pebble Bed Type High Temperature Gas-cooled Reactor with annular core structure

    International Nuclear Information System (INIS)

    This report presents the Conceptual Design Study of Pebble Bed Type High Temperature Gas-cooled Reactor with Annular Core Structure. From this study, it is made clear that the thermal power of the Pebble Bed Type Reactor can be increased to 500MW through introducing the annular core structure without losing the inherent safe characteristics (in the coolant depressurization accident, the fuel temperature does not exceed the temperature where the fuel defect begins.) This thermal power is two times higher than the inherent safe Pebble Bed Type High temperature Gas-cooled Reactor (MHTGR) designed in West Germany. From this result, it is foreseen that the ratio of the plant cost to the reactor power is reduced and the economy of the plant operation is improved. The reactor performances e.g. fuel burnup and fuel temperature are maintained in same level of the MHTGR. (author)

  19. 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.

  20. Dynamics and transient stability of a pebble bed reactor during start up

    Energy Technology Data Exchange (ETDEWEB)

    Miles, B.; Pain, C.C.; Eaton, M.D.; Ziver, A.K.; Goddard, A.J.H. [Applied Modelling and Computation Group, Imperial College of Science, Technology and Medicine, Dept. of Earth Science and Engineering, London (United Kingdom); Oliveira, C.R.E. de [Nuclear and Radiological Engineering and Medical Physics Program, The George W Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA (United States)

    2005-07-01

    A design of a modular pebble bed reactor (PBR) is being developed for construction in South Africa. The design of this PBR is simulated in the FETCH nuclear criticality model. FETCH solves the neutron transport equations coupled to fluid dynamics and has been used in simulations of fluidized bed reactors. In the neutronics module of FETCH steady state neutronic calculations are performed to obtain the starting conditions for the subsequent calculation of transient behaviour. These include fuel temperature and control rod position. Neutron flux and the initial surplus reactivity are also calculated. Each step change in a simulated start-up is initiated by an excess reactivity which produces more severe transients than would be encountered in normal operation. The variations of several parameters with time are recorded, for example, temperature at various points in the reactor, temperature of the hottest pebble and fission rate. Spatial profiles are recorded at regular time intervals, including temperatures, power density, gas velocity and gas pressure. The stability of the reactor is demonstrated.

  1. Modular pebble-bed reactor reforming plant design for process heat

    International Nuclear Information System (INIS)

    This report describes a preliminary design study of a Modular Pebble-Bed Reactor System Reforming (MPB-R) Plant. The system uses one pressure vessel for the reactor and a second pressure vessel for the components, i.e., reformer, steam generator and coolant circulator. The two vessels are connected by coaxial pipes in an arrangement known as the side-by-side (SBS). The goal of the study is to gain an understanding of this particular system and to identify any technical issues that must be resolved for its application to a modular reformer plant. The basic conditions for the MPB-R were selected in common with those of the current study of the MRS-R in-line prismatic fuel concept, specifically, the module core power of 250 MWt, average core power density of 4.1 w/cc, low enriched uranium (LEU) fuel with a 235U content of 20% homogeneously mixed with thorium, and a target burnup of 80,000 MWD/MT. Study results include the pebble-bed core neutronics and thermal-hydraulic calculations. Core characteristics for both the once-through-then-out (OTTO) and recirculation of fuel sphere refueling schemes were developed. The plant heat balance was calculated with 55% of core power allotted to the reformer

  2. A Safe Solution to World Energy Supply - the Very High Temperature Pebble Bed Reactor

    International Nuclear Information System (INIS)

    For the energy hungry world there is a solution which has the potential to resolve most of the present energy needs, with almost zero pollution and high thermal efficiency. The Very High Temperature Reactor (VHTR) can produce Hydrogen for automotive needs to replace the polluting gas and oil; it can produce electricity at very high efficiency with almost no pollution, and provide clean process heat for the industry and the energy needed for desalination plants to provide fresh water. In the present study it is shown that choosing the Pebble Bed concept for the VHTR is not only a very effective way to supply all the energy needs, it is also one of the safest nuclear reactor concept. Depending on the fuel cycle chosen, it is possible to reduce significantly the TRU waste normally produced in light water reactors and thus further reduce the environmental concerns of long living FP. A conceptual 600MWt High Temperature Pebble Bed reactor is proposed, and its safety characteristics are analyzed by simulating various hypothetical accidents, using the DSNP simulation system

  3. Simplified models for pebble-bed HTR core burn-up calculations with Monteburns2.0©

    International Nuclear Information System (INIS)

    Highlights: ► PBMR-400 annular core is very difficult to simulate in a reliable way. ► Nuclide evolutions given by different lattice models can differ significantly. ► To split fixed lattice models into two axial zones does not affect results significantly. ► We can choose a (simplified) core model on the basis of the analysis aim. ► Monteburns gives by survey burn-up calculations reasonable nuclide evolution trends. - Abstract: This paper aims at comparing some simplified models to simulate irradiation cycles of Pu fuelled pebble-bed reactors with Monteburns2.0© code. As a reference core, the PBMR-400 (proposed in the framework of the EU PUMA project, where this kind of core fuelled by a Pu and Pu–Np fuel has been studied) was taken into account. Pebble-bed High Temperature Reactor (HTR) cores consist of hundreds of thousands pebbles arranged stochastically in a cylindrical or annular space and each pebble is a single fuel element, and it is able to reach ultra-high burn-ups, i.e. up to 750 GWd/tHM (for Pu-based fuels). Additionally, pebble-bed cores are characterised by a continuous recirculation of pebbles from the top to the bottom of the core. Modelling accurately with current computer codes such an arrangement, in order to predict the behaviour of the core itself, is a very difficult task and any depletion code specifically devoted to pebble-bed burn-up calculation is not available at the moment. Because of limitations of the most common current MCNP-based depletion codes as well as huge calculation times, simplified models have to be implemented. After an analysis of the literature available on pebble-bed models for criticality and burn-up calculations, a preliminary assessment of the impact of different kind of simplified models for a Pu-Np fuelled Pebble-Bed Modular Reactor (PBMR), proposed in the framework of the EU PUMA project, is shown, particularly as far as burn-up prediction with Monteburns2.0© code is concerned.

  4. Autonomous multi-purpose floating power system with a compact static pebble bed reactor

    International Nuclear Information System (INIS)

    The paper introduces a new concept of an autonomous multipurpose system with a compact static-bed pebble bed reactor as a power source. The system is envisioned as a small floating power complex in which a compact high-efficiency nuclear system provides the source of energy for a variety of industrial processes. It offers the near-term (with a conventional power source) and long-term (with a compact high-efficiency nuclear system) technologies for a low cost electricity/potable water supply compared to traditional systems for regions where local communities are isolated and do not have extensive industrial infrastructure and distribution networks. The complex can be quickly installed anywhere following demands and needs of local communities - coastal regions and islands. The reactor design and system layout, balance-of-plant evaluations, performance characteristics and deployment strategies are discussed. (authors)

  5. Autonomous multi-purpose floating power system with a compact static pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsvetkov, Pavel; Vierow, Karen; Peddicord, Kenneth; Ragusa, Jean; McDeavitt, Sean; Poston, John Sr.; Shao, Lin; Willems, Greg [Department of Nuclear Engineering, Texas A and M University, College Station, Texas (United States)

    2008-07-01

    The paper introduces a new concept of an autonomous multipurpose system with a compact static-bed pebble bed reactor as a power source. The system is envisioned as a small floating power complex in which a compact high-efficiency nuclear system provides the source of energy for a variety of industrial processes. It offers the near-term (with a conventional power source) and long-term (with a compact high-efficiency nuclear system) technologies for a low cost electricity/potable water supply compared to traditional systems for regions where local communities are isolated and do not have extensive industrial infrastructure and distribution networks. The complex can be quickly installed anywhere following demands and needs of local communities - coastal regions and islands. The reactor design and system layout, balance-of-plant evaluations, performance characteristics and deployment strategies are discussed. (authors)

  6. Supplemental Report on Nuclear Safeguards Considerations for the Pebble Bed Modular Reactor (PBMR)

    International Nuclear Information System (INIS)

    Recent reports by Department of Energy National Laboratories have discussed safeguards considerations for the low enriched uranium (LEU) fueled Pebble Bed Modular Reactor (PBMR) and the need for bulk accountancy of the plutonium in used fuel. These reports fail to account effectively for the degree of plutonium dilution in the graphitized-carbon pebbles that is sufficient to meet the International Atomic Energy Agency's (IAEA's) '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 may not be judged sufficiently diluted to meet the IAEA provisional guidelines for termination of safeguards, the estimated quantities of 232U and 236U in the used fuel at the target burn-up of ∼91 GWD/MT exceed specification limits for reprocessed uranium (ASTM C787) and will require extensive blending with either natural uranium or uranium enrichment tails to dilute the 236U content to fall within specification thus making the PBMR used fuel less desirable for commercial reprocessing and reuse than that from light water reactors. Also the PBMR specific activity of reprocessed uranium isotopic mixture and its A2 values for effective dose limit if released in a dispersible form during a transportation accident are more limiting than the equivalent values for light water reactor spent fuel at 55 GWD/MT without accounting for the presence of the principal carry-over fission product (99Tc) and any possible plutonium contamination that may be present from attempted covert reprocessing. Thus, the potentially recoverable uranium from PBMR 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

  7. Characteristics of convective heat transport in a packed pebble-bed reactor

    International Nuclear Information System (INIS)

    Highlights: • A fast-response heat transfer probe has been developed and used in this work. • Heat transport has been quantified in terms of local heat transfer coefficients. • The method of the electrically heated single sphere in packing has been applied. • The heat transfer coefficient increases from the center to the wall of packed bed. • This work advancing the knowledge of heat transport in the studied packed bed. - Abstract: Obtaining more precise results and a better understanding of the heat transport mechanism in the dynamic core of packed pebble-bed reactors is needed because this mechanism poses extreme challenges to the reliable design and efficient operation of these reactors. This mechanism can be quantified in terms of a solid-to-gas convective heat transfer coefficient. Therefore, in this work, the local convective heat transfer coefficients and their radial profiles were measured experimentally in a separate effect pilot-plant scale and cold-flow experimental setup of 0.3 m in diameter, using a sophisticated noninvasive heat transfer probe of spherical type. The effect of gas velocity on the heat transfer coefficient was investigated over a wide range of Reynolds numbers of practical importance. The experimental investigations of this work include various radial locations along the height of the bed. It was found that an increase in coolant gas flow velocity causes an increase in the heat transfer coefficient and that effect of the gas flow rate varies from laminar to turbulent flow regimes at all radial positions of the studied packed pebble-bed reactor. The results show that the local heat transfer coefficient increases from the bed center to the wall due to the change in the bed structure, and hence, in the flow pattern of the coolant gas. The findings clearly indicate that one value of an overall heat transfer coefficient cannot represent the local heat transfer coefficients within the bed; therefore, correlations are needed to

  8. On the evaluation of pebble bed reactor critical experiments using the PEBBED code

    International Nuclear Information System (INIS)

    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 assumptions, however, is shown to be minimal. The embedded transport solver can also be used to obtain control rod worths but only with adjustment of the local spectrum. Results are compared to those of other codes as well as Core 4 of the HTR-Proteus experiment which contains partially inserted rods. They indicate the need for a reference solution to adjust the radius of the graphite in the

  9. On the evaluation of pebble bed reactor critical experiments using the PEBBED code

    Energy Technology Data Exchange (ETDEWEB)

    Hans D. Gougar; R. Sonat Sen

    2001-10-01

    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 assumptions, however, is shown to be minimal. The embedded transport solver can also be used to obtain control rod worths but only with adjustment of the local spectrum. Results are compared to those of other codes as well as Core 4 of the HTR-Proteus experiment which contains partially inserted rods. They indicate the need for a reference solution to adjust the radius of the graphite in the

  10. High temperature gas-cooled pebble bed reactor steady state thermal-hydraulics analyses based on CFD method

    International Nuclear Information System (INIS)

    Background: Based on general purpose CFD code Fluent, the PBMR-400 full load nominal condition thermal-hydraulics performance was studied by applying local thermal non-equilibrium porous media model. Purpose: In thermal hydraulics study of the gas cooled pebble bed reactor, the core of the reactor can be treated as macroscopic porous media with strong inner heat source, and the original Fluent code can not handle it properly. Methods: By introducing a UDS in the calculation domain of the reactor core and subjoining a new resistance term, we develop a non-equilibrium porous media model which can give an accurate description of the core of the pebble bed. The mesh of CFD code is finer than that of the traditional pebble bed reactor thermal hydraulics analysis code such as THERMIX and TINTE, thus more information about coolant velocity fields, temperature field and solid phase temperature field can be acquired. Results: The nominal condition calculation results of the CFD code are compared to those of the well-established thermal-hydraulic code THERMIX and TINTE, and show a good consistency. Conclusion: The extended local thermal non-equilibrium model can be used to analyse thermal-hydraulics of high temperature pebble bed type reactor. (authors)

  11. Plutonium and minor actinide utilisation in a pebble-bed high temperature reactor

    International Nuclear Information System (INIS)

    This paper contains results of the analysis of the pebble-bed high temperature gas-cooled PUMA reactor loaded with plutonium and minor actinide (Pu/MA) fuel. Starting from knowledge and experience gained in the Euratom FP5 projects HTR-N and HTR-N1, this study aims at demonstrating the potential of high temperature reactors to utilize or transmute Pu/MA fuel. The work has been performed within the Euratom FP6 project PUMA. A number of different fuel types and fuel configurations have been analyzed and compared with respect to incineration performance and safety-related reactor parameters. The results show the excellent plutonium and minor actinide burning capabilities of the high temperature reactor. The largest degree of incineration is attained in the case of an HTR fuelled by pure plutonium fuel as it remains critical at very deep burnup of the discharged pebbles. Addition of minor actinides to the fuel leads to decrease of the achievable discharge burnup and therefore smaller fraction of actinides incinerated during reactor operation. The inert-matrix fuel design improves the transmutation performance of the reactor, while the 'wallpaper' fuel does not have advantage over the standard fuel design in this respect. After 100 years of decay following the fuel discharge, the total amount of actinides remains almost unchanged for all of the fuel types considered. Among the plutonium isotopes, only the amount of Pu-241 is reduced significantly due to its relatively short half-life. (authors)

  12. Plutonium and minor actinide utilisation in a pebble-bed high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Petrov, B. Y.; Kuijper, J. C.; Oppe, J.; De Haas, J. B. M. [Nuclear Research and Consultancy Group, Westerduinweg 3, 1755 ZG Petten (Netherlands)

    2012-07-01

    This paper contains results of the analysis of the pebble-bed high temperature gas-cooled PUMA reactor loaded with plutonium and minor actinide (Pu/MA) fuel. Starting from knowledge and experience gained in the Euratom FP5 projects HTR-N and HTR-N1, this study aims at demonstrating the potential of high temperature reactors to utilize or transmute Pu/MA fuel. The work has been performed within the Euratom FP6 project PUMA. A number of different fuel types and fuel configurations have been analyzed and compared with respect to incineration performance and safety-related reactor parameters. The results show the excellent plutonium and minor actinide burning capabilities of the high temperature reactor. The largest degree of incineration is attained in the case of an HTR fuelled by pure plutonium fuel as it remains critical at very deep burnup of the discharged pebbles. Addition of minor actinides to the fuel leads to decrease of the achievable discharge burnup and therefore smaller fraction of actinides incinerated during reactor operation. The inert-matrix fuel design improves the transmutation performance of the reactor, while the 'wallpaper' fuel does not have advantage over the standard fuel design in this respect. After 100 years of decay following the fuel discharge, the total amount of actinides remains almost unchanged for all of the fuel types considered. Among the plutonium isotopes, only the amount of Pu-241 is reduced significantly due to its relatively short half-life. (authors)

  13. 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.

  14. Proceedings of the fifteenth international workshop on ceramic breeder blanket interactions

    International Nuclear Information System (INIS)

    This report is the Proceedings of 'the Fifteenth International Workshop on Ceramic Breeder Blanket Interactions' which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors. This workshop was held in Sapporo, Japan on 3-4, Sept. 2009. Twenty six participants from EU, Japan, India, Russia and USA attended the workshop. The scope of the workshop included 1) evolutions in ceramic breeder blanket design, 2) progress in ceramic breeder material development, 3) irradiation testing, 4) breeder material properties, 5) out-of-pile pebble bed experiment, 6) modeling of the thermal, mechanical and tritium transfer behavior of pebble beds and 7) interfacing issues of solid breeder blanket development. By this workshop, advance of key technologies for solid breeder blanket development was shared among the participants. Also, desired direction of further investigation and development was recognized. The 20 of the presented papers are indexed individually. (J.P.N.)

  15. 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

  16. Pebble Bed Reactor Plant screening evaluation. Volume 1. Overall plant and reactor system

    International Nuclear Information System (INIS)

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW/sub t/ Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system. Core scoping studies were performed which evaluated the effects of annular and cyclindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations

  17. Advanced pebble bed high temperature reactor with central graphite column for future applications

    International Nuclear Information System (INIS)

    Design evaluations of the advanced pebble bed high temperature reactor, AHTR, with central graphite column are given. This reactor, as a nuclear heat source, is suitable for coal refinement as well as for electricity generation with closed gas turbine primary helium circuit. With this design of the central graphite column, it is possible to limit the core temperatures under the required value of about 1600deg C in case of accident conditions, even with higher thermal power and higher core inlet and outlet temperatures. The designs of core internals are described. The after heat removal system is integrated in the prestressed concrete reactor pressure vessel, which is based on the principals of natural convection. Research work is being carried out, whereby the sphencal fuel elements are coated with a layer of silicon carbide, to improve the corrosion resistance as well as the effectiveness of the fission products barrier. (orig.)

  18. DSNP models used in the pebble-bed HTGR dynamic simulation. V.2

    International Nuclear Information System (INIS)

    A detailed description is given of the components that were used in the DSNP simulation of the PNP-500 high temperature gas-cooled pebble-bed reactor. Each component presented in this report describes in detail the mathematical model that was used, and the assumptions that were made in developing the model. Most of the models were developed using basic physical principles with the simplication that could be justified on the basis of the requested accuracy. Most of the models were developed as either one dimensional or lumped parameter models. The heat transfer and flow correlations, which are mostly based on semiempirical correlations were either provided by KFA or were adapted from the available literature. A short description of DSNP is also given, with a comprehensive list of all the statements available in Rev. 4.1 of DSNP. (H.K.)

  19. Neutronic behavior of Thorium based fuel cycles in a pebble bed reactor

    International Nuclear Information System (INIS)

    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. This paper shows the main advantages of the use of a Pebble Bed critical nuclear reactor using a variety of fuel cycles with Thorium (Th+U233, Th+Pu239 and Th+U). the parameters related to the neutronic behavior like deep burn, nuclear fuel breeding, Minor Actinide stockpile, power density profiles and other are used to compare the fuel cycles. also a thermo mechanical study of the irradiated TRISO fuel particle is presented. (Author)

  20. Fission product release out of the core of a pebble bed reactor in core heatup accidents

    International Nuclear Information System (INIS)

    This report presents the analysis of fission product release from the core of a pebble-bed high temperature reactor during hypothetical accidents. First the models describing fission product transport are discussed, and on the basis of these models a computer code is developped. This code includes the diffusion of fission products from particles and through the graphite, and the sorption of metallic fission product elements on graphite as well as the plateout of metallic fission product elements in the top- and bottom reflectors. In addition a review of the necessary empirical input data is given. Then the cesium release of a single fuel element at high temperatures is calculated, and the results are compared with experimental data. Furthermore calculations of the fission product release from the core of a 500 MW(th) high temperature reactor during core heatup accidents are made, and the influence of the most important parameters is described. (orig.)

  1. Effect of packing fraction variations on reactivity in pebble-bed reactor

    International Nuclear Information System (INIS)

    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)

  2. Process for fuelling a pebble bed reactor and device for carrying out this process

    International Nuclear Information System (INIS)

    In a process for fuelling the inside of the pebble bed reactor with pellet fuel elements, these are supplied from above through a gravity device and are distributed radially, where the gravity device has a central gravity tube. So that the expense of the device, particularly the number of gravity tubes can be reduced, all fuel elements are introduced into the inside of the reactor only through the central gravity tube and the speed of fall of the fuel elements is varied, so that when they impinge on the surface of the cone which forms during fuelling, very probably they either remain in the centre or are deflected radially, corresponding to the selected speed of fall. (orig.)

  3. Pyro-hydrolysis treatment technology of the ion exchange resin by pebble-bed pyrolysis reactor

    International Nuclear Information System (INIS)

    Treatment method of spent ion-exchange resin and sludge for sub surface disposal, such as from reactor water clean-up system and from fuel pool cooling cleanup system, is unestablished even now. However, such technology is essential task because increasing the volume of these wastes from decommissioning is unavoidable problem. Due to the solutions, NGK INSULATORS, LTD (NGK) has successfully developed and confirmed that pyro-hydrolysis can treat these waste. This treatment technology, which is based on pebble-bed pyrolyser in nitrogen atmosphere by NUKEM Technologies GmbH, has been improved with superheated steam. This report describes that overview, characteristic and examination result of the treatment technology for such resins with pyro-hydrolysis. (author)

  4. Storage built pebble bed for greenhouse use; Acumulador tipo lecho para uso en invernaderos

    Energy Technology Data Exchange (ETDEWEB)

    Bistoni, S.; Iriarte, A.; Saravia, L.

    2004-07-01

    To heat greenhouses during the night it is necessary to use storage systems. Our region shows high radiation levels, even in winter, so during the day stored energy inside of a greenhouse is more than necessary. If the excess of energy is stored up, it can be used during the night when it is necessary. In this paper the performance of a storage built with plastic bottles with water inside them are studied and it is compared with a pebble bed. A model and its solution using the electric- thermal analogy are presented. The results show that it is feasible and economic to build a storage like the proposed. It is important to mention that the simulation is very simple because of the computational program used. (Author)

  5. Gas reactor international cooperative program interim report: German Pebble Bed Reactor design and technology review

    International Nuclear Information System (INIS)

    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

  6. 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.

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

    International Nuclear Information System (INIS)

    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 9500C, 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 9500C 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

  8. Development of a thermal–hydraulic analysis code for the Pebble Bed Water-cooled Reactor

    International Nuclear Information System (INIS)

    Highlights: ► Main design features of the PBWR were put forward. ► Thermal–hydraullics analysis code for the PBWR was developed and verified. ► Key thermal–hydraullics parameters were calculated in normal operation. ► The PBWR has a great pressure loss but an excellent heat transfer characteristic. ► Maximum fuel temperature and MDNBR are in conformity with safety criterion. - Abstract: The Pebble Bed Water-cooled Reactor (PBWR) is a water-moderated water-cooled pebble bed reactor in which millions of tristructural-isotropic (TRISO) coated micro-fuel elements (MFE) pile in each assembly. Light water is used as coolant that flows from bottom to top in the assembly while the moderator water flows in the reverse direction out of the assembly. Steady-state thermal–hydraullic analysis code for the PBWR will provide a set of thermal hydraulic parameters of the primary loop so that heat transported out of the core can match with the heat generated by the core for a safe operation of the reactor. The key parameters of the core including the void fraction, pressure drop, heat transfer coefficients, the temperature distribution and the Departure from Nucleate Boiling Ratio (DNBR) is calculated for the core in normal operation. The code can calculate for liquid region, water-steam two phase region and superheated steam region. The results show that the maximum fuel temperature is much lower than the design limitation and the flow distribution can meet the cooling requirement in the reactor core. As a new type of nuclear reactor, the main design features with a sufficient safety margin were also put forward in this paper.

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

    International Nuclear Information System (INIS)

    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 keff value compared to VSOP and PEBBED, respectively. While using JENDL-3.3, MCPBR gave a 2.22% and 3.11% higher keff 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

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

    International Nuclear Information System (INIS)

    The AVR pebble bed reactor (46 MWth) 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 MWth). 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

  11. 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

  12. Progress in the development of Li{sub 2}ZrO{sub 3} and Li{sub 2}TiO{sub 3} pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Lulewicz, J.D.; Roux, N. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France)

    1998-03-01

    Li{sub 2}ZrO{sub 3} and Li{sub 2}TiO{sub 3} pebbles are being developed as ceramic breeder for the European Helium-cooled pebble bed DEMO blanket concept. Status is given of the fabrication work, and of the properties characteristics determination. (author)

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

    International Nuclear Information System (INIS)

    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

  14. Thermal cycling tests on Li4SiO4 and beryllium pebbles

    International Nuclear Information System (INIS)

    The European B.O.T. Demo-relevant solid breeder blanket is based on the use of beds of beryllium and Li4SiO4 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 Li4SiO4 pebbles. No breaking was observed in the beryllium pebbles, however the Li4SiO4 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.)

  15. Supplemental Report on Nuclear Safeguards Considerations for the Pebble Bed Modular Reactor (PBMR)

    Energy Technology Data Exchange (ETDEWEB)

    Moses, David Lewis [ORNL; Ehinger, Michael H [ORNL

    2010-05-01

    Recent reports by Department of Energy National Laboratories have discussed safeguards considerations for the low enriched uranium (LEU) fueled Pebble Bed Modular Reactor (PBMR) and the need for bulk accountancy of the plutonium in used fuel. These reports fail to account effectively for the degree of plutonium dilution in the graphitized-carbon pebbles that is sufficient to meet the International Atomic Energy Agency's (IAEA's) '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 may not be judged sufficiently diluted to meet the IAEA provisional guidelines for termination of safeguards, the estimated quantities of {sup 232}U and {sup 236}U in the used fuel at the target burn-up of {approx}91 GWD/MT exceed specification limits for reprocessed uranium (ASTM C787) and will require extensive blending with either natural uranium or uranium enrichment tails to dilute the {sup 236}U content to fall within specification thus making the PBMR used fuel less desirable for commercial reprocessing and reuse than that from light water reactors. Also the PBMR specific activity of reprocessed uranium isotopic mixture and its A{sub 2} values for effective dose limit if released in a dispersible form during a transportation accident are more limiting than the equivalent values for light water reactor spent fuel at 55 GWD/MT without accounting for the presence of the principal carry-over fission product ({sup 99}Tc) and any possible plutonium contamination that may be present from attempted covert reprocessing. Thus, the potentially recoverable uranium from PBMR 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

  16. Special topics reports for the reference tandem mirror fusion breeder: liquid metal MHD pressure drop effects in the packed bed blanket. Vol. 1

    International Nuclear Information System (INIS)

    Magnetohydrodynamic (MHD) effects which result from the use of liquid metal coolants in magnetic fusion reactors include the modification of flow profiles (including the suppression of turbulence) and increases in the primary loop pressure drop and the hydrostatic pressure at the first wall of the blanket. In the reference fission-suppressed tandem mirror fusion breeder design concept, flow profile modification is a relatively minor concern, but the MHD pressure drop in flowing the liquid lithium coolant through an annular packed bed of beryllium/thorium pebbles is directly related to the required first wall structure thickness. As such, it is a major concern which directly impacts fissile breeding efficiency. Consequently, an improved model for the packed bed pressure drop has been developed. By considering spacial averages of electric fields, currents, and fluid flow velocities the general equations have been reduced to simple expressions for the pressure drop. The averaging approach results in expressions for the pressure drop involving a constant which reflects details of the flow around the pebbles. Such details are difficult to assess analytically, and the constant may eventually have to be evaluated by experiment. However, an energy approach has been used in this study to bound the possible values of the constant, and thus the pressure drop. In anticipation that an experimental facility might be established to evaluate the undetermined constant as well as to address other uncertainties, a survey of existing facilities is presented

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

    Directory of Open Access Journals (Sweden)

    Zhu Guifeng

    2016-01-01

    Full Text Available 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-BeF2 salt Temperature Reactivity Coefficient (TRC. Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tristructural-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 good safety margin. The greatest challenge of this reactor may be the decades irradiation time of the pebble fuel.

  18. Fuel Management Optimization of Pebble-Bed Reactors Using Particle Swarm Algorithm

    International Nuclear Information System (INIS)

    A procedure for the optimization of pebble bed reactors fuel management, utilizing PSO algorithm has been developed. This procedure has been used for optimizing the natural uranium utilization of a large 3000MWth core, operating with an OTTO fuel management scheme. Low enriched uranium and thorium fuel cycles have been investigated. Thorium was assumed to be loaded in the form of mixed Th-U oxide fuel and also in separate Th pebbles. The optimization results indicate that thorium introduction does not improve natural uranium utilization when constraining uranium enrichment to non-proliferation level of 20%, however it does decrease Pu production by 50% compared to the low enriched uranium cycle. Only 3% improvement in natural uranium utilization is gained with mixed oxide of highly enriched uranium and thorium fuel. In this case, Pu production is reduced by 95%. When introducing Th to the fuel, higher uranium enrichments are required, up to the 20% when constrained by the proliferation limit or more than 70% when unconstrained. This enriched uranium demand increases natural uranium requirements for the enrichment process, which is compensated by reduced overall core uranium loading. Optimized Th introduction also results in longer fuel residence time at to core, for the efficient buildup of U233. Th introduction also effects moderation ratio, from high value of 550 for LEU case, it reduces down to 497 and 461for the proliferation constrained Th-MOX and SEP cases and down to 297 for the unconstrained Th- MOX case respectively. Th-MOX and SEP cases may have lower fuel costs by up to 17%. Fuel management with fuel loading in two zones has only slightly improved performance since in this large core, the radial flux distribution is already relatively uniform. MEDUL fuel cycle axial distribution resembles the cosine shape with reduced power density and hence maximum fuel temperatures

  19. 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)

  20. 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

  1. Characterization of constrained beryllium pebble beds after neutron irradiation at HFR at high temperatures up to helium production of 3000 appm

    Energy Technology Data Exchange (ETDEWEB)

    Chakin, V., E-mail: vladimir.chakin@kit.edu [Institute for Applied Materials – Applied Materials Physics, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Plarz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rolli, R. [Institute for Applied Materials – Materials and Biomechanics, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Plarz 1, 76344 Eggenstein-Leopoldshafen (Germany); Moeslang, A.; Vladimirov, P.; Kurinskiy, P. [Institute for Applied Materials – Applied Materials Physics, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Plarz 1, 76344 Eggenstein-Leopoldshafen (Germany); Til, S. van; Magielsen, A.J. [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/ Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain)

    2013-10-15

    Highlights: • Defragmentation of beryllium pebbles at irradiation temperatures of 873 and 948 K was detected. • Formation of brittle beryllium oxide layers on neutron irradiated beryllium pebbles was detected. • Strong interaction between beryllium pebbles and platinum foil under neutron irradiation was detected. • Strong interaction between beryllium pebbles and austenitic stainless steel under neutron irradiation was detected. -- Abstract: Small constrained beryllium pebble beds as well as unconstrained beryllium pebbles have been irradiated within HIDOBE-01 experiment at HFR, Petten, the Netherlands. Beryllium pebbles with 1 mm diameter produced by Rotating Electrode Method (REM) were investigated after irradiation at 630, 740, 873, and 948 K up to helium production of 3000 appm. Intensive pore and bubble formation occurs in beryllium after 873 K irradiation. In the contact zones of the pebbles enhanced pore formation takes place. Oxidation of beryllium pebble external surfaces is accompanied by partial destruction of oxide layers owing to their high brittleness. Strong interactions between beryllium pebbles and platinum foil, as well as between beryllium and stainless steel at contact zones occur at 873 and 948 K.

  2. Status and perspective of the R and D on ceramic breeder materials for testing in ITER

    International Nuclear Information System (INIS)

    The main line of ceramic breeder materials research and development is based on the use of the breeder material in the form of pebble beds. At present, there are three candidate pebble materials (Li4SiO4, and two forms of Li2TiO3) for DEMO reactors that will be used for testing in ITER. This paper reviews the R and D of as-fabricated pebble materials against the blanket performance requirements and makes recommendations on necessary steps toward the qualification of these materials for testing in ITER

  3. Experimental study of bypass flow in near wall gaps of a pebble bed reactor using hot wire anemometry technique

    International Nuclear Information System (INIS)

    Highlights: • Coolant flow behavior in near wall gaps of a pebble bed reactor is studied. • Hot wire anemometry is applied for high frequency velocity measurements. • Bypass flow is identified within the velocity profiles of near wall gaps. • Effect of gap geometry and Reynolds number on bypass flow is investigated. • Variation of velocity power spectra with radial location and Reynolds number is studied. - Abstract: Coolant flow behavior through the core of an annular pebble bed reactor is investigated in this experimental study. A high frequency hot wire anemometry system coupled with an X-probe is used for measurement of axial and radial velocity components at different points within two near wall gaps at five different modified Reynolds numbers (Rem = 2043–6857). The velocity profiles within the gaps verify the presence of an area of increased velocity close to the pebble bed outer reflector wall, which is known as the bypass flow. Moreover, the characteristics of the coolant flow profile are seen to be highly dependent on the gap geometry. The effect of Reynolds number on the velocity profiles varies as the geometry of the gap changes. The time histories of the local velocities measured with considerably high frequency are further analyzed using power spectral density technique. Power spectral plots illustrate substantial spatial variation of the energy content, spectral shape, and the slope of the energy cascade region. A significant correlation between Reynolds number and characteristics of the velocity power spectra is observed

  4. Effective Thermal Property Estimation of Unitary Pebble Beds Based on a CFD-DEM Coupled Method for a Fusion Blanket

    Science.gov (United States)

    Chen, Lei; Chen, Youhua; Huang, Kai; Liu, Songlin

    2015-12-01

    Lithium ceramic pebble beds have been considered in the solid blanket design for fusion reactors. To characterize the fusion solid blanket thermal performance, studies of the effective thermal properties, i.e. the effective thermal conductivity and heat transfer coefficient, of the pebble beds are necessary. In this paper, a 3D computational fluid dynamics discrete element method (CFD-DEM) coupled numerical model was proposed to simulate heat transfer and thereby estimate the effective thermal properties. The DEM was applied to produce a geometric topology of a prototypical blanket pebble bed by directly simulating the contact state of each individual particle using basic interaction laws. Based on this geometric topology, a CFD model was built to analyze the temperature distribution and obtain the effective thermal properties. The current numerical model was shown to be in good agreement with the existing experimental data for effective thermal conductivity available in the literature. supported by National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2015GB108002, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

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

    International Nuclear Information System (INIS)

    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. Pebble bed modular reactor - The first Generation IV reactor to be constructed

    Energy Technology Data Exchange (ETDEWEB)

    Ion, S. [British Nuclear Fuels plc, Warrington (United Kingdom); Nicholls, D. [ESKOM, Sandton, Johannesburg (South Africa); Matzie, R. [Westinghouse Electric Company, Windsor, CT (United States); Matzner, D. [Pebble Bed Modular Reactor (Pty) Ltd, Centurion (South Africa)

    2004-02-01

    Substantial interest has been generated in advanced reactors over the past few years. This interest is motivated by the view that new nuclear power reactors will be needed to provide low carbon generation of electricity and possibly hydrogen to support the future growth in demand for both of these commodities. Some governments feel that substantially different designs will be needed to satisfy the desires for public perception, improved safety, proliferation resistance, reduced waste and competitive economics. This has motivated the creation of the Generation IV Nuclear Energy Systems programme in which ten countries have agreed on a framework for international cooperation in research for advanced reactors. Six designs have been selected for continued evaluation, with the objective of deployment by 2030. One of these designs is the very high temperature reactor (VHTR), which is a thermal neutron spectrum system with a helium-cooled core utilising carbon-based fuel. The pebble bed modular reactor (PBMR), being developed in South Africa through a worldwide international collaborative effort led by Eskom, the national utility, will represent a key milestone on the way to achievement of the VHTR design objectives, but in the much nearer term. This paper outlines the design objectives, safety approach and design details of the PBMR, which is already at a very advanced stage of development. (author)

  7. Pebble bed modular reactors versus other generation technologies. Costs and challenges for South Africa

    International Nuclear Information System (INIS)

    South Africa is Africa's major economy, with plans to double its electricity generation capacity by 2026. South Africa has spent almost two decades developing a nuclear reactor known as a Pebble Bed Modular Reactor (PBMR), which could provide substantial benefits to the electricity grid but was recently mothballed due to high costs. This work estimates the lifecycle financial costs of South African PBMRs, then compares these costs to those of five other generation options: coal, nuclear as pressurized water reactors (PWRs), wind, and solar as photovoltaics (PV) or concentrating solar power (CSP). Each technology is evaluated with low, base case, and high assumptions for capital costs, construction time, and interest rates. Decommissioning costs, project lifetime, capacity factors, and sensitivity to carbon price are also considered. PBMR could be cost competitive with coal under certain low cost conditions, even without a carbon price. However, international lending practices and other factors suggest that a high capital cost, high interest rate nuclear plant is likely to be competing with a low capital cost, low interest rate coal plant in a market where cost recovery is challenging. PBMR could potentially become more competitive if low rate international loans were available to nuclear projects or became unavailable to coal projects. (author)

  8. Acacia: A small scale power plant with pebble bed cartridge reactor and indirect Brayton cycle

    International Nuclear Information System (INIS)

    For markets other than large-scale electricity production a 60 MWth, 23 MWe (max.) nuclear plant design with an indirect Brayton cycle is proposed for application on the short to medium term. The reactor will be cooled by helium, whereas for the secondary cycle nitrogen is proposed as a heat carrier. In this way, a conventional air based gas turbine can be applied, while at the same time excluding the scenario of air ingress in the reactor core through a heat exchanger leak. Two variations of cycle design will be discussed: co-generation and maximized electricity production. The cogeneration mode will be elaborated for the application of seawater desalination. The reactor core geometry is annular with a central graphite reflector region, creating an optimal location for burnable poison. Optimization calculations on burnable poison distribution show that burnup of fuel and poison are balancing each other into a fairly constant reactivity behaviour during the entire core lifetime. Also, the two most important safety transient scenarios for pebble bed reactors, Pressurised and Depressurised Loss Of Forced Cooling, will be discussed. It will be shown that the maximum fuel temperatures will stay below the level where fuel damage starts for any point in time. (author)

  9. 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.

  10. Simulation of the pebble bed modular reactor natural air convection passive heat removal system

    International Nuclear Information System (INIS)

    Cooling of the Pebble Bed Nuclear Reactor under evaluation in South Africa is primarily effected by the flow of helium through the cavity which contains the nuclear fuel. However, apart from this, a certain amount of heat flows from the reactor cavity, through the graphite barrel and reactor vessel to the containment building and ultimately to the environment During normal operation this passive heat loss represents approximately 1MW for a 100MW reactor and constitutes an undesirable loss of power. In the event of a shutdown or loss of main coolant, however, this passive heat removal is relied upon to remove the decay heat from the core. A study was initiated to simulate the process of this heat removal to provide an indication of the maximum vessel temperature and power transfers after shutdown. However, there is a lack of precise data indicating values for thermal conductivity, heat transfer coefficients, heat capacities or even densities. This paper describes the assumptions made and the manner in which these data were estimated so as to provide what is hoped to be a reasonably accurate estimate of the behaviour of the passive heat removal process. (author)

  11. A preliminary conceptual design for a 150 MWth pebble bed reactor core using the VSOP94 code package

    International Nuclear Information System (INIS)

    Recently, the hydrogen production using heat source of the high-temperature gas-cooled reactor (HTGR) has been attracting worldwide attention. Since the domestic neutronic design codes for the LWR core design were judged not to be applicable to design of the HTGRs, the VSOP94 code system for the core designs of pebble bed type HTGRs was installed in order to be used in the HTGR design until the domestic code system would be developed. After the VSOP94 was verified against a benchmark calculation for the PRROTEUS experiment, the preliminary conceptual design for a hypothetical bed reator with 150 MWth was performed using the VSOP94

  12. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    Science.gov (United States)

    Scarlat, Raluca Olga

    This dissertation treats system design, modeling of transient system response, and characterization of individual phenomena and demonstrates a framework for integration of these three activities early in the design process of a complex engineered system. A system analysis framework for prioritization of experiments, modeling, and development of detailed design is proposed. Two fundamental topics in thermal-hydraulics are discussed, which illustrate the integration of modeling and experimentation with nuclear reactor design and safety analysis: thermal-hydraulic modeling of heat generating pebble bed cores, and scaled experiments for natural circulation heat removal with Boussinesq liquids. The case studies used in this dissertation are derived from the design and safety analysis of a pebble bed fluoride salt cooled high temperature nuclear reactor (PB-FHR), currently under development in the United States at the university and national laboratories level. In the context of the phenomena identification and ranking table (PIRT) methodology, new tools and approaches are proposed and demonstrated here, which are specifically relevant to technology in the early stages of development, and to analysis of passive safety features. A system decomposition approach is proposed. Definition of system functional requirements complements identification and compilation of the current knowledge base for the behavior of the system. Two new graphical tools are developed for ranking of phenomena importance: a phenomena ranking map, and a phenomena identification and ranking matrix (PIRM). The functional requirements established through this methodology were used for the design and optimization of the reactor core, and for the transient analysis and design of the passive natural circulation driven decay heat removal system for the PB-FHR. A numerical modeling approach for heat-generating porous media, with multi-dimensional fluid flow is presented. The application of this modeling

  13. Conceptual Design Studies of a Passively Safe Thorium Breeder Pebble Bed Reactor

    OpenAIRE

    Wols, F.J.

    2015-01-01

    Nuclear power plants are expected to play an important role in the worldwide electricity production in the coming decades, since they provide an economically attractive, reliable and low-carbon source of electricity with plenty of resources available for at least the coming hundreds of years. However, the design of nuclear reactors can be improved significantly in terms of safety, by designing reactors with fully passive safety systems, and sustainability, by making more efficient use of natu...

  14. Conceptual Design Studies of a Passively Safe Thorium Breeder Pebble Bed Reactor

    NARCIS (Netherlands)

    Wols, F.J.

    2015-01-01

    Nuclear power plants are expected to play an important role in the worldwide electricity production in the coming decades, since they provide an economically attractive, reliable and low-carbon source of electricity with plenty of resources available for at least the coming hundreds of years. Howeve

  15. Li ceramic pebbles chemical compatibility with Eurofer samples in fusion relevant conditions

    International Nuclear Information System (INIS)

    Information on the chemical compatibility between Li ceramic breeders and reactor structural materials is an important issue for fusion reactor technology. In this work, Eurofer samples were placed inside a Li ceramic pebble bed and kept at 600 deg. C under a reducing atmosphere obtained by the flow of a purging gas (He + 0.1vol.%H2). Titanate and orthosilicate Li pebble beds were used in the experiments and exposure time ranged from 50 to 2000 h. Surface chemical reactions were investigated with nuclear microprobe techniques. The orthosilicate pebbles present chemical reactions even with the gas mixture, whereas for the samples in close contact with Eurofer there is evidence of Eurofer elemental diffusion into the pebbles and the formation of different types of compounds. Although the titanate pebbles used in the chemical compatibility experiments present surface alterations with increasing surface irregularities along the annealing time, there is no clear indication of Eurofer constituents diffusion

  16. Nuclear analyses for two 'look-alike' helium-cooled pebble bed test blanket sub-modules proposed by the US for testing in ITER

    International Nuclear Information System (INIS)

    The US is proposing two 'look-alike' sub-modules, based on helium-cooled pebble bed (HCPB) ceramic breeder, to be tested in the same test blanket module (TBM) that will occupy a quarter of a port in ITER and placed next to the Japanese TBM. The TBM has a toroidal width of 73 cm, a radial depth of 60 cm and a poloidal height of 91 cm. The ceramic breeder is made of Li4SiO4 with 75% Li-6 enrichment (60% packing factor) and beryllium is used as the multiplier. The two sub-modules are arranged in two configurations, namely a layered configuration and an edge-on configuration. In the present work, we analyze these two sub-modules using two-dimensional discrete ordinates transport codes in R-θ model that accounts for the presence of the ITER shielding blanket and the surrounding frame of the port. The objectives are: (1) to examine the profiles of heating and tritium production rates in the two sub-modules, both in the radial and toroidal direction, in order to identify locations where neutronics measurements can be best performed with least perturbation from the surroundings (2) to provide both local and integrated values for nuclear heating rates required for subsequent thermo-mechanics analysis, and (3) to compare the tritium production capabilities of two variants for the HCPB blanket concept, mainly the parallel and the edge-on configurations. We present the main findings from this study in this paper

  17. Transient modelling of sulphur-iodine cycle thermochemical hydrogen generation coupled to pebble bed modular reactor

    International Nuclear Information System (INIS)

    A transient control volume model of the sulphur iodine (S-I) and Westinghouse hybrid sulphur (HyS) cycles is presented. These cycles are some of the leading candidates for hydrogen generation using a high temperature heat source. The control volume models presented here are based on a heat and mass balance in each reaction chamber coupled to the relevant reaction kinetics. The chemical kinetics expressions are extracted from a relevant literature review. Two assumptions regarding reaction chamber pressure are identified, namely a constant pressure condition and a differential form of ideal gas law. The HyS model is based on an application of the Nernst equation. This application of the Nernst equation suggests that in the HyS cycle the hydrogen generation rate is directly proportional to the SO2 production rate. The observed chemical kinetic response time of the sulphuric acid decomposition section is on the order of 30 seconds, whereas the response time of the hydrogen iodide decomposition section is on the order of 500 seconds. It is concluded that the decomposition of hydrogen iodide (HI) is the rate limiting step of the entire S-I cycle. High temperature nuclear reactors are ideal candidates for use as a driving heat source for both the S-I and HyS cycle. The pebble bed modular reactor is a type of very high temperature reactor (VHTR ) suitable for nuclear hydrogen generation. A methodology for coupling of the S-I or HyS cycle to a pebble bed modular reactor (PBMR) via an intermediate heat exchanger (IHX) is developed. A 2-D THERMIX heat transfer model of a PBMR-268 is presented, and this model is coupled to a point kinetics model. The point kinetics model was developed to meet the same specifications as the RELAP5 point kinetics module. A steady-state integration of the S-I and HyS cycle models to the PBMR 268 heat transfer model is performed. The integration assumes that 100% of the heat energy from the PBMR-268 is deposited into the chemical plant via the

  18. Approach to development of high flux research reactor with pebble-bed core

    International Nuclear Information System (INIS)

    Full text: The research nuclear reactor of a basin-type IRT with the designed power of 1 MW was put into operation in 'Sosny' settlement not far from Minsk-city in the Republic of Belarus in 1962. In 1971 after its modernization the power was increased up to 4 MW and maximum density of neutron flux in the core was: Thermal 5·1013 neutr./cm2.s Fast (E>0.8 MeV) 2·1013 neutr./cm2.s The reactor has been used for carrying out investigations in the field of solid-state physics, radiation construction materials, radiobiology, gaseous chemically reacting coolants and others. After the Chernobyl NPP accident, in the former USSR the requirements on safety of nuclear reactors have become sufficiently stricter. As to some parameters these requirements became the same as for reactors of nuclear power plants. In this connection the reactor in 'Sosny' settlement did not answer these new requirements by a number of performances such as seismicity of building, efficiency of control and protection system, corrosion in the reactor vessel and others, and it was shutdown in 1987 and its decommissioning was performed during 1988-1999. At the Joint Institute of Power and Nuclear Research -'SOSNY' have been carried out investigations on feasibility of creation of the research reactor with pebble-bed core. The concept of such reactor supposes using the following technical approaches: - Using as fuel the brought sphere micro fuel elements with the diameter of 500-750 mkm to an industrial level; - Organization of reactor operation in the regime with minimum possible fueling with 235U; - Implementation of hydraulic loading - unloading of micro fuel elements with the frequency of one or several days. Physical calculations of the core were carried out with the help of MCU-RFFI program based on the Monte-Carlo method. Two configurations of the pebble-bed core in the high flux reactor have been considered. The first configuration is the core with a neutron trap and an annular fuel layer formed

  19. Core and fuel design for Pebble Bed Modular Reactor (PBMR) using SRAC computer code

    International Nuclear Information System (INIS)

    Core and fuel down scale analysis on PBMR-HTR using SRAC program aims to identify the influence of U235 enrichment, burnable poison, coolant flow rate and coolant temperature entered to criticality core and safety aspects of nuclear reactor with the parameters are multiplication factor (keff) and fuel temperature coefficient, moderator temperature coefficient and coolant temperature coefficient. Core PBMR-HTR finite cylindrical with a hole in the middle which contains 334,000 pebble fuel bed. That consist of UO2 fuel, graphite moderator and helium coolant. Down scale the design model performed on the half core represent the whole core. The study was conducted by varying the fuel enrichment of 8%; 8.5%; 9%; 9.5% and 10%, while variation burnable poison enrichment at 5 ppm, 7 ppm, 9 ppm, 11 ppm and 15 ppm. The variation of coolant flow rate of 60%, 80%, 100%, 120% and 140% from its original value at 17.118 kg/s while the variation of coolant temperature input at 673.15 K; 723.15 K; 773.15 K; 823.15 K and 873.15 K. In this research, value of keff without Gd2O3 are 1.026213 (BOL) and 1.004173 (EOL) with excess reactivity of 2.55% with 9% U235 enrichment. While keff on BOL by using 7 ppm Gd2O3 of 1.006968 and 1.004198 for EOL with excess reactivity of 0.69%. Fuel temperature reactivity coefficient, moderator and coolant in a row for -8.597317E-05/K; -2.595284E-05 /K and 1.1496E-06/K. Temperature reactivity coefficient is negative. This indicates inherent safety characteristic have been met. Increasing the input temperature and coolant flow rate reduction lowers the value of keff core, and it will contribute to negative reactivity coefficient. (author)

  20. PEBBLES Mechanics Simulation Speedup

    Energy Technology Data Exchange (ETDEWEB)

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2010-05-01

    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. These simulations involve hundreds of thousands of pebbles and involve determining the entire core motion as pebbles are recirculated. Single processor algorithms for this are insufficient since they would take decades to centuries of wall-clock time. This paper describes the process of parallelizing and speeding up the PEBBLES pebble mechanics simulation code. Both shared memory programming with the Open Multi-Processing API and distributed memory programming with the Message Passing Interface API are used in simultaneously in this process. A new shared memory lock-less linear time collision detection algorithm is described. This method allows faster detection of pebbles in contact than generic methods. These combine to make full recirculations on AVR sized reactors possible in months of wall clock time.

  1. Examination of the potential for diversion or clandestine dual use of a pebble-bed reactor to produce plutonium

    International Nuclear Information System (INIS)

    This paper explores the susceptibility of Pebble-Bed Reactors (PBRs) to be used overtly or covertly for the production of plutonium for nuclear weapons. The basic assumption made for the consideration of overt production is that a country would purchase a PBR with the ostensible motive of producing electric power; then, after the power plant was built, the country would divert the facility entirely to the production of weapons material. It is assumed that the country would then have to manufacture production pebbles from natural uranium. The basic assumption made for covert production is that the country would obtain and use a PBR for power production, but that it would clandestinely feed plutonium production pebbles through the reactor in such small numbers that the perturbation on power plant operation would be very difficult to detect. This paper shows the potential rate of plutonium production under such constraints. It is demonstrated that the PBR is a very poor choice for either form of proliferation-intent use. (author)

  2. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 2. Conceptual balance of plant design

    International Nuclear Information System (INIS)

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. This volume describes the conceptual balance-of-plant (BOP) design and was prepared by United Engineers and Constructors, Inc. of Philadelphia, Pennsylvania. The major emphasis of the BOP study was a preliminary design of an overall plant to provide a basis for future studies

  3. Parametric study of thorium fuel cycles in a 100MWth pebble bed high temperature reactor / F. Panday

    OpenAIRE

    Panday, Farisha

    2011-01-01

    The current project was conducted in order to select an optimized open Thorium/Uranium fuel cycle for the Pebble Bed Modular Reactor (PBMR) concept in motivation for the 100MWth PBMR Power Plant. A sensitivity study on the heavy metal loading of the fuel sphere was performed to accomplish this task. The effect on various parameters was evaluated to determine the influence of varying the Heavy Metal (HM) from 6 gHM/sphere to 20 gHM/sphere and at different feed fuel enrichment...

  4. On the numerical assessment of the thermo-mechanical performances of the DEMO Helium-Cooled Pebble Bed breeding blanket module

    International Nuclear Information System (INIS)

    Highlights: • HCPB blanket module thermo-mechanical behavior has been investigated under normal operation and over-pressurization steady state scenarios. • A theoretical–computational approach based on the finite element method (FEM) has been followed, adopting a qualified commercial FEM code. • Under normal operation scenario, SDC-IC safety rule relevant to the loss of ductility is not fulfilled in the FW and in the hot spots of SPv. • Under over-pressurization scenario, SDC-IC safety rule relevant to the loss of ductility is not met in the hot spots of lower and upper SPv. - Abstract: Within the framework of the European DEMO Breeder Blanket Programme, a research campaign has been launched by University of Palermo, ENEA-Brasimone and Karlsruhe Institute of Technology to theoretically investigate the thermo-mechanical behavior of the Helium-Cooled Pebble Bed (HCPB) breeding blanket module of the DEMO1 blanket vertical segment, under normal operation and over-pressurization loading scenarios. The research campaign has been carried out following a theoretical–computational approach based on the finite element method (FEM) and adopting a qualified commercial FEM code. A realistic 3D FEM model of the HCPB blanket module central poloidal–radial region has been developed, including one breeder cell in the toroidal direction and all the five cells in the poloidal one. No Breeder Units have been modeled, their presence being simulated by effective thermo-mechanical loads. Two sets of uncoupled steady state thermo-mechanical analyses have been carried out with reference to the investigated loading scenarios. In particular, under normal operation scenario (level A) the module has been supposed to undergo both 8 MPa coolant pressure on its cooling channel walls and thermal deformations due to the flat-top plasma operational state thermal field, while under over-pressurization scenario (level D) it has been assumed to experience 8 MPa coolant pressure on its

  5. From field to factory-Taking advantage of shop manufacturing for the pebble bed modular reactor

    International Nuclear Information System (INIS)

    The move of nuclear plant construction from the field to the factory for small, advanced pebble bed modular reactor (PBMR) designs has significant benefits compared to traditional light water reactor (LWR) field oriented designs. The use of modular factory construction techniques has a growing economic benefit over time through well-established process learning applications. This paper addresses the basic PBMR design objectives and commercialization model that drive this approach; provides a brief technical description of the PBMR design and layout with representative CAD views and discusses derived figures of merit highlighting the relative simplicity of PBMR compared to a modern LWR. The discussion emphasizes that more of PBMR can be built in the factory due to the simple design of a direct helium Brayton cycle compared to an indirect LWR steam cycle with its associated equipment. For the PBMR design there are fewer and less cumbersome auxiliary and safety systems with their attendant support requirements. Additionally, the labor force economic efficiency for nuclear projects is better in the factory than in the field, including consideration of labor costs and nuclear quality programs. Industrial learning is better in the factory because of the more controlled environment, mechanization optimization opportunities and because of the more stable labor force compared to the field. Supply chain benefits are more readily achievable with strategic contracts for module suppliers. Although building a nuclear power plant is not a typical high volume manufacturing process, for the PBMR-type of plant, with its high degree of standardization and relatively small, simplified design, the shift to factory work has a significant impact on overall project cost due to earlier identification and better coordination of parallel construction paths. This is in stark contrast to the construction of a large LWR in the past. Finally, the PBMR modular plant concept continues at the

  6. Progress in the Development of the Modular Pebble-Bed Advanced High Temperature Reactor

    International Nuclear Information System (INIS)

    This review article summarizes recent progress by students and faculty at U.C. Berkeley working on the development of the Pebble-Bed Advanced High Temperature Reactor (PB-AHTR). The 410-MWe PBAHTR is a liquid salt cooled reactor that operates at near atmospheric pressure and high power density (20 to 30 MW/m3, compared to 4.8 MW/m3 for helium cooled reactors). Operating with a core inlet temperature of 600 deg. C and outlet temperature of 704 deg. C, the PB-AHTR uses well understood materials of construction including Alloy 800H with Hastelloy N cladding for the reactor vessel and primary loop components, and graphite for core and reflector structures. Recent work by the NE 170 senior design class has developed physical arrangements for the major reactor and power conversion components, along with the structural design for the reactor building and turbine hall featuring seismic base isolation, design for aircraft crash protection, shielding analysis, and design of a multiple-zone ventilation and containment system to provide effective control of radioactive and chemical contamination. The resulting total building volume is 260 m3/MWe, compared to 343 m3/MWe to 486 m3/MWe for current large (1150 to 1600 MWe) LWR designs. These results suggest the potential for significant reductions in construction time and cost. Neutronics studies have verified the capability to design the PB-AHTR with negative fuel and coolant temperature reactivity coefficients, for both LEU and deep-burn TRU fuels. Depletion analysis was also performed to identify optimal core designs to maximize fuel utilization. The additional moderation provided by the coolant simplifies design to achieve optimal moderation, and the spent fuel volume is approximately half that of helium cooled reactors. In collaboration with the Czech Nuclear Research Institute, initial zero-power critical tests were performed to validate PB-AHTR neutronics models. Liquid salts are unique among candidate reactor coolants due

  7. Temperature cycling tests on a mixed Be/Li4SiO4 pebble bed in the HEBLO facility. Final report

    International Nuclear Information System (INIS)

    The second HEBLO experiment with a mixed pebble bed corresponding to the forerunner concept of the helium cooled breeding blanket for DEMO has been completed successfully. The experiment was conducted to simulate a DEMO-related cyclic load of the pebble bed. The pebble bed survived the entire series of experiments totaling 1915 cycles under a variety of different loads without suffering any major damage. The result of subsequent examination was in good agreement with the qualitative evaluation of the temperatures measured and the pressure losses measured in the purge gas. Recalculations of the experiment performed in accordance with the DEMO design principles showed good agreement of the transient temperatures with the measured levels. (orig.)

  8. Test apparatus for ITER blanket pebble packing behavior

    International Nuclear Information System (INIS)

    Current Japanese design for ITER Driver Blanket consists of three breeder layers, nine multiplier layers and five cooling panels. The breeder layers and the multiplier layers contain 1 mm diameter spheres of Li2O and Be, respectively. The heat transfer in such 'Pebble Layered Blanket' is largely affected by the packing fraction of the pebbles which can be easily changed by the vibration during the operation. The packing fraction of the pebbles are expected to be as high as possible on the view point of nuclear heat design to maintain the optimum temperature of the breeder layer. Thus, it is necessary to establish the stable packed bed of the breeder and multiplier. The present experimental apparatus was fabricated for the engineering tests with the partial model of Japanese blanket. Test apparatus consists of stainless steel test panels, transparent plastic test panels, vibrators and measurement instruments. The apparatus can examine various parameters of sphere packed beds such as packing fraction, panels deformation, loading weight at the bottom of the panels and so on under various vibrating conditions. (author)

  9. Non-classical particle transport with angular-dependent path-length distributions. II: Application to pebble bed reactor cores

    International Nuclear Information System (INIS)

    Highlights: • We construct and analyze random and crystal arrangements of pebble bed cores. • We investigate anisotropic diffusion of neutrons in the interior of the cores. • We generate benchmark numerical results using Monte Carlo. • We obtain model estimates for the anisotropic diffusion coefficients. • We find that the new theory is accurate and able to predict anisotropic diffusion. - Abstract: We describe an analysis of neutron transport in the interior of model pebble bed reactor (PBR) cores, considering both crystal and random pebble arrangements. Monte Carlo codes were developed for (i) generating random realizations of the model PBR core, and (ii) performing neutron transport inside the crystal and random heterogeneous cores; numerical results are presented for two different choices of material parameters. These numerical results are used to investigate the anisotropic behavior of neutrons in each case and to assess the accuracy of estimates for the diffusion coefficients obtained with the diffusion approximations of different models: the atomic mix model, the Behrens correction, the Lieberoth correction, the generalized linear Boltzmann equation (GLBE), and the new GLBE with angular-dependent path-length distributions. This new theory utilizes a non-classical form of the Boltzmann equation in which the locations of the scattering centers in the system are correlated and the distance-to-collision is not exponentially distributed; this leads to an anisotropic diffusion equation. We show that the results predicted using the new GLBE theory are extremely accurate, correctly identifying the anisotropic diffusion in each case and greatly outperforming the other models for the case of random systems

  10. Theoretical and experimental research of natural convection in the core of the high temperature pebble bed reactor

    International Nuclear Information System (INIS)

    The physical model of the developed THERMIX-2D-code for computing thermohydraulic behaviour of the core of high temperature pebble bed reactors is verified by experiments with natural convection flow. Such fluid flow behaviour can be of very high importance for the real reactor in the case of natural heat removal decay. The experiments are performed in a special set up testing-stand with pressures up to 30 bars and temperatures up to 3000C by using air and helium as fluid. In comparison with the experimental data the numerical results show that a good and useful simulation is given by the program. Pure natural convection flow in packed pebble beds is calculated with a very high degree of reliability. The investigation of flow stability demonstrate that radial-symmetric relations are not given temporarily when national convection is overlayed by forced convection flow. In the discussion it is explained when and to what extent the program leds to useful results in such situations. The test of the effective heat conductivity lambdasub(eff) results in an improvement of the lambdasub(eff)-data used so far for temperatures below 13000C. (orig.)

  11. Fluid flow and heat transfer investigation of pebble bed reactors using mesh adaptive large-eddy simulation

    International Nuclear Information System (INIS)

    A computational fluid dynamics model with anisotropic mesh adaptivity is used to investigate coolant flow and heat transfer in pebble bed reactors. A novel method for implicitly incorporating solid boundaries based on multi-fluid flow modelling is adopted. The resulting model is able to resolve and simulate flow and heat transfer in randomly packed beds, regardless of the actual geometry, starting off with arbitrarily coarse meshes. The model is initially evaluated using an orderly stacked square channel of channel-height-to-particle diameter ratio of unity for a range of Reynolds numbers. The model is then applied to the face-centred cubical geometry. Coolant flow and heat transfer patterns are investigated. (author)

  12. Tritium release from lithium orthosilicate pebbles deposited with palladium

    International Nuclear Information System (INIS)

    Full text of publication follows: Slightly over-stoichiometric lithium orthosilicate pebbles have been selected as one optional breeder material for the European Helium Cooled Pebble Bed (HCPB) blanket. This material has been developed in collaboration of Research Center Karlsruhe and the Schott Glass, Mainz. The lithium orthosilicate pebbles are fabricated from lithium hydroxide and silica by a melting and spraying method in a semi-industrial scale facility. Lithium hydroxide was selected as the precursor since enriched lithium hydroxide is commercially available. The lithium orthosilicate pebbles produced by the process contains oxide phases besides orthosilicate, but it was also found that the oxide phases can be decomposed by annealing at high temperatures. The lithium orthosilicate pebbles produced in this way possesses satisfactory pebble characteristics. Therefore, the authors performed out-of-pile annealing tests using the lithium orthosilicate pebbles irradiated in a research reactor. Moreover, the effect of the deposition of palladium in the lithium orthosilicate pebbles on the behavior of tritium release was investigated. Palladium was deposited in the lithium orthosilicate pebbles by the incipient wet impregnation method using a solution of a palladium amino complex. The lithium orthosilicate pebbles were submitted to neutron irradiation at the Kyoto university research reactor. In the out-of-pile annealing experiments, the temperature of the breeder material placed in a tubular reactor made of quartz was raised from ambient temperature to 1173 K at a constant rate of 5 K/min under the stream of sweep gases. The tritium concentration in the outlet stream of the reactor was traced with two ionization chambers. The ionization chambers were used with a water bubbler, which enables to measure the concentrations of molecular form of tritium (HT) and tritiated water vapor (HTO) separately. In the experiments, a 0.1 % hydrogen/nitrogen sweep gas was used. The

  13. Cynod: A Neutronics Code for Pebble Bed Modular Reactor Coupled Transient Analysis

    International Nuclear Information System (INIS)

    The Pebble Bed Reactor (PBR) is one of the two concepts currently considered for development into the Next Generation Nuclear Plant (NGNP). This interest is due, in particular, to the concept's inherent safety characteristics. In order to verify and confirm the design safety characteristics of the PBR computational tools must be developed that treat the range of phenomena that are expected to be important for this type of reactors. This paper presents a recently developed 2D R-Z cylindrical nodal kinetics code and shows some of its capabilities by applying it to a set of known and relevant benchmarks. The new code has been coupled to the thermal hydraulics code THERMIX/KONVEK(1) for application to the simulation of very fast transients in PBRs. The new code, CYNOD, has been written starting with a fixed source solver extracted from the nodal cylindrical geometry solver contained within the PEBBED code. The fixed source solver was then incorporated into a kinetic solver. The new code inherits the spatial solver characteristics of the nodal solver within PEBBED. Thus, the time-dependent neutron diffusion equation expressed analytically in each node of the R-Z cylindrical geometry sub-domain (or node) is transformed into one-dimensional equations by means of the usual transverse integration procedure. The one-dimensional diffusion equations in each of the directions are then solved using the analytic Green's function method. The resulting equations for the entire domain are then re-cast in the form of the Direct Coarse Mesh Finite Difference (D-CMFD) for convenience of solution. The implicit Euler method is used for the time variable discretization. In order to correctly treat the cusping effect for nodes that contain a partially inserted control rod a method is used that takes advantage of the Green's function solution available in the intrinsic method. In this corrected treatment, the nodes are re-homogenized using axial flux shapes reconstructed based on the Green

  14. Design modeling of fuel particles for high-burnup in pebble-bed fast reactors

    International Nuclear Information System (INIS)

    The thermomechanical and neutron lifetime of different fuel particle designs is assessed by applying a new performance modeling platform comprised of an analytical stress code and finite element engineering hydrocode. Our investigation is based on fuel for fast reactors with the goal of high-burnup to provide minimal waste disposal. Fuel designs are considered based on variations of the standard Modular Pebble-Bed Reactor (MPBR) design in which the spherical fuel kernel contained by a three-layer coating system comprised of inner Pyrolytic Carbon (IPyC), Silicon Carbide (SiC), and outer PyC (OPyC). The neutronics calculations used in our investigation are based on a new fusion-fission engine concept called LIFE (Laser Inertial Confinement Fusion-Fission Energy). Particle stresses are calculated accounting for the interplay between mechanisms such as irradiation-induced swelling and creep, thermal expansion, anisotropic elastic effects, and layer a sphericity. In addition, mechanisms such as corrosion and void coalescence are considered in order to avoid failure of the particles by way of layer cracking and leakage of the fission products, or other pathways. Our design investigation involves a parametric study of layer materials with respect to their thermal conductivity, irradiation resistance, constitutive and other properties and layer thickness to develop a fuel particle design with optimized resistance to failure mechanisms for the desired operating conditions. A key component of the modeling platform is the capability to examine the time and space evolution of all mechanisms affecting performance which are often neglected for the conditions at low burn-up levels. Specifically, temperature variation as a function of depth into the layers generates stresses and also affects the amount of swelling, particularly at high fluence. Moreover, irradiation temperature cycling has been identified as a source of additional time-varying stresses that can lead to cracking

  15. Optimization of coupled multiphysics methodology for safety analysis of pebble bed modular reactor

    Science.gov (United States)

    Mkhabela, Peter Tshepo

    The research conducted within the framework of this PhD thesis is devoted to the high-fidelity multi-physics (based on neutronics/thermal-hydraulics coupling) analysis of Pebble Bed Modular Reactor (PBMR), which is a High Temperature Reactor (HTR). The Next Generation Nuclear Plant (NGNP) will be a HTR design. The core design and safety analysis methods are considerably less developed and mature for HTR analysis than those currently used for Light Water Reactors (LWRs). Compared to LWRs, the HTR transient analysis is more demanding since it requires proper treatment of both slower and much longer transients (of time scale in hours and days) and fast and short transients (of time scale in minutes and seconds). There is limited operation and experimental data available for HTRs for validation of coupled multi-physics methodologies. This PhD work developed and verified reliable high fidelity coupled multi-physics models subsequently implemented in robust, efficient, and accurate computational tools to analyse the neutronics and thermal-hydraulic behaviour for design optimization and safety evaluation of PBMR concept The study provided a contribution to a greater accuracy of neutronics calculations by including the feedback from thermal hydraulics driven temperature calculation and various multi-physics effects that can influence it. Consideration of the feedback due to the influence of leakage was taken into account by development and implementation of improved buckling feedback models. Modifications were made in the calculation procedure to ensure that the xenon depletion models were accurate for proper interpolation from cross section tables. To achieve this, the NEM/THERMIX coupled code system was developed to create the system that is efficient and stable over the duration of transient calculations that last over several tens of hours. Another achievement of the PhD thesis was development and demonstration of full-physics, three-dimensional safety analysis

  16. 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.

  17. 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.

  18. 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

    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.

  19. One-dimensional modeling of radial heat removal during depressurized heatup transients in modular pebble-bed and prismatic high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    A one-dimensional computational model was developed to evaluate the heat removal capabilities of both prismatic-core and pebble-bed modular HTGRs during depressurized heatup transients. A correlation was incorporated to calculate the temperature- and neutron-fluence-dependent thermal conductivity of graphite. The modified Zehner-Schluender model was used to determine the effective thermal conductivity of a pebble bed, accounting for both conduction and radiation. Studies were performed for prismatic-core and pebble-bed modular HTGRs, and the results were compared to analyses performed by GA and GR, respectively. For the particular modular reactor design studied, the prismatic HTGR peak temperature was 2152.20C at 38 hours following the transient initiation, and the pebble-bed peak temperature was 1647.80C at 26 hours. These results compared favorably with those of GA and GE, with only slight differences caused by neglecting axial heat transfer in a one-dimensional radial model. This study found that the magnitude of the initial power density had a greater effect on the temperature excursion than did the initial temperature

  20. Gaseous and metallic fission product release characteristics of a modular pebble bed HTGR during loss of core cooling accidents

    International Nuclear Information System (INIS)

    A quantitative safety criteria for the high-temperature gas-cooled reactor (HTGR) is to limit the radiological consequences for a wide spectrum of accidents to a level not requiring public sheltering. This leads to reliance on passive safety characteristics for improbable loss of core cooling accidents. Models have been developed to predict the transport of metallic and gaseous fission products (FPs) through the multilayered fuel particle coatings and the graphite matrix of the core under accident conditions. Using these models, FP transport and releases were calculated for a loss of core convective cooling accident in a 250-MW(t) 3.8-W/cc pebble bed HTGR. Fission-product transport through the particle kernel and coatings, the graphite pebbles/reflectors, the reactor vessel, and the confinement were assessed. The results of this study show that the most effective barrier to fission products is the coated fuel particle. The reactor vessel and the confinement provide additional attenuation for the small amount released from the core. The small release to the environment occurs over a period of days and is so low that the safety criterion of 5 rem thyroid dose (to avoid offsite sheltering) is satisfied with a margin of more than an order of magnitude. 6 figs

  1. Core design of NPP Pebble Bed Modular Reactor (PBMR) type using computer code MCNP-5 for beginning of life (BOL)

    International Nuclear Information System (INIS)

    The core design of Nuclear Power Plant for Pebble Bed Modular Reactor (PBMR) type with 70 MWe capacity power in Beginning of Life (BOL) has been performed. The aim of this analysis, to know percent enrichment, temperature distribution and safety value by negative temperature coefficient at type PBMR if reactor power become lower equal to 70 MWe. This analysis was expected become one part of overview project development the power plant with 10.000 MWe of total capacity, spread evenly in territory of Indonesia especially to support of smelter industries. The results showed that, effective multiplication factor (keff) with power 70 MWe critical condition at enrichment 5,626 % is 1,00031±0, 00087, based on enrichment result, a value of the temperature coefficient reactivity is 10,0006 pcm/K. Based on the results of these studies, it can be concluded that the PBMR 70 MWe design is theoretically safe. (author)

  2. The modular pebble bed nuclear reactor - the preferred new sustainable energy source for electricity, hydrogen and potable water production?

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  4. Methods For The Calculation Of Pebble Bed High Temperature Reactors With High Burnup Plutonium And Minor Actinide Based Fuel

    International Nuclear Information System (INIS)

    The graphite moderated Modular High Temperature Pebble Bed Reactor enables very flexible loading strategies and is one candidate of the Generation IV reactors. For this reactor fuel cycles with high burnup (about 600 MWd/kg HM) based on plutonium (Pu) and minor actinides (MA) fuel will be investigated. The composition of this fuel is defined in the EU-PuMA-project which aims the reduction of high level waste. There exist nearly no neutronic full core calculations for this fuel composition with high burnup. Two methods (deterministic and Monte Carlo) will be used to determine the neutronics in a full core. The detailed results will be compared with respect to the influence on criticality and safety related parameters. (authors)

  5. Adaptation of the South African regulatory framework to the licensing of the pebble bed modular reactor. Regulatory challenges

    International Nuclear Information System (INIS)

    Internationally, it has been recognized that there is a need to adapt the regulatory systems and regulations in the countries being faced with the introduction of new nuclear technologies and applications, thus posing some challenges to the regulatory framework of such countries. With the development of the pebble bed modular reactor, being pursued by South Africa as one of its alternative energy sources, the South African regulatory framework and licensing philosophy had to be adapted in terms of ensuring that a credible and effective licensing process be developed and implemented for this 'new' technology. This paper will present the major challenges which the South African National Nuclear Regulator faced in developing and implementing such a licensing process and how these are being addressed. The paper will also discuss the stakeholders' involvement and interaction in this project as required by the relevant South African legislation. (author)

  6. Loss-of-water accident analysis of the pebble-bed modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The high pressure helium and water/steam are respectively used as the primary and secondary coolant for the pebble-bed modular high temperature gas-cooled reactor (HTGR). Loss-of-water accident is one of the typical design basis accident (DBA), which would be caused by malfunction or current failure of the feed water pump, as well as the false action of the feed water valve. During the loss-of-water accident, due to the loss of the secondary heat sink, the temperature and pressure of primary coolant will increase. Subsequently, the reactor scram will be triggered by the protective signal of the “high flow rate proportion of primary circuit and secondary circuit” or the “high core inlet helium temperature”. For this type of the accident, the earlier open of the safety valve of the primary circuit should be avoided by reactor design. Based on the preliminary design of the 250 MW pebble-bed modular high temperature gas-cooled reactor (HTR-PM), with the coupled analysis code TINTE-BLAST, accidents with different slowdown rate of the feed water supply have been studied. The important parameters, including the reactor power, fuel element temperature, inlet/outlet helium temperature of the core, and especially the primary pressure, are analyzed. The consequences with first scram signal succeeding or failing are compared. The results can prove that, according to the current design of the protection system, this kind of accident can be detected in time. The scram signal will trigger the reactor shut down quickly, without causing the earlier open of the safety valve. After the reactor is successfully shut down, due to the inherent safety feature of the HTGR, the temperature and the pressure in the primary circuit will increase very slowly. The temperature of the fuel element, as well as that of the components, will not exceed the design limitations. (author)

  7. Coated fuel particle temperature analysis of the pebble-bed modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    In the 200 MWe pebble-bed modular high temperature gas-cooled reactor (HTR-PM), each sphere fuel element contains approximately 12,000 coated fuel particles scattered in the inner graphite matrix with a diameter of 50 mm to form the fuel zone, while the outer shell with a thickness of 5 mm is a fuel-free zone made up of the same graphite material. The coated fuel particle, with a diameter of less than 1 mm, consists of a UO2 kernel in 0.5 mm diameter and multiple coated ceramic layers. The HTR-PM has good inherent safety properties, one of which is exhibited like that, under some transient or accidental situations leading to an unexpected power increase, the reactor can shut down itself automatically or be brought down to a very low power level only by the negative temperature coefficient of reactivity due to the fuel temperature rise. During the calculation of the fuel element temperature with the pebble bed reactor analysis software THERMIX, which was originally developed by the German KFA-Juelich, a uniform power density in the fuel zone is assumed, without considering the temperature difference between the coated fuel particles and the surrounding graphite matrix. In this paper, the reactor temperature feedback characteristics and the nuclear power during a rapid reactivity introduction accident are analyzed in detail for two cases, i.e. taking into account the coated fuel particle temperature or not. The calculation results show that, the coated fuel particle temperature rises more quickly than the graphite matrix, and then the reactor power descends after a limited increase due to the higher negative temperature coefficient of reactivity of the fuel particle compared with that of the graphite moderator. Besides, the calculation conservation of the THERMIX code is revealed, and the safety properties of the HTR-PM are illustrated as well. (authors)

  8. Corrosion susceptibility of EUROFER97 in lithium ceramics breeders

    International Nuclear Information System (INIS)

    EUROFER97 specimens were exposed in vacuum to lithium silicate pebbles at 550 °C for up to 2880 h, to evaluate its corrosion susceptibility in a simulated breeder blanket environment. The specimens and pebble bed were then analyzed and characterized by SEM-EDX, XRD, and HR-TEM. The results revealed the formation of a double chromium/iron oxide corrosion layer. HR-TEM also showed that the inner layer was amorphous, while the outer was crystalline. The amorphous layer was brittle, broke easily, and became detached from the steel

  9. 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.

  10. Comparison among MCNP-based depletion codes applied to burnup calculations of pebble-bed HTR lattices

    International Nuclear Information System (INIS)

    The double-heterogeneity characterising pebble-bed high temperature reactors (HTRs) makes Monte Carlo based calculation tools the most suitable for detailed core analyses. These codes can be successfully used to predict the isotopic evolution during irradiation of the fuel of this kind of cores. At the moment, there are many computational systems based on MCNP that are available for performing depletion calculation. All these systems use MCNP to supply problem dependent fluxes and/or microscopic cross sections to the depletion module. This latter then calculates the isotopic evolution of the fuel resolving Bateman's equations. In this paper, a comparative analysis of three different MCNP-based depletion codes is performed: Montburns2.0, MCNPX2.6.0 and BGCore. Monteburns code can be considered as the reference code for HTR calculations, since it has been already verified during HTR-N and HTR-N1 EU project. All calculations have been performed on a reference model representing an infinite lattice of thorium-plutonium fuelled pebbles. The evolution of k-inf as a function of burnup has been compared, as well as the inventory of the important actinides. The k-inf comparison among the codes shows a good agreement during the entire burnup history with the maximum difference lower than 1%. The actinide inventory prediction agrees well. However significant discrepancy in Am and Cm concentrations calculated by MCNPX as compared to those of Monteburns and BGCore has been observed. This is mainly due to different Am-241 (n,γ) branching ratio utilized by the codes. The important advantage of BGCore is its significantly lower execution time required to perform considered depletion calculations. While providing reasonably accurate results BGCore runs depletion problem about two times faster than Monteburns and two to five times faster than MCNPX.

  11. Criticality calculations on pebble-bed HTR-PROTEUS configuration as a validation for the pseudo-scattering tracking method implemented in the MORET 5 Monte Carlo code

    International Nuclear Information System (INIS)

    The MORET code is a three dimensional Monte Carlo criticality code. It is designed to calculate the effective multiplication factor (keff) of any geometrical configuration as well as the reaction rates in the various volumes and the neutron leakage out of the system. A recent development for the MORET code consists of the implementation of an alternate neutron tracking method, known as the pseudo-scattering tracking method. This method has been successfully implemented in the MORET code and its performances have been tested by mean of an extensive parametric study on very simple geometrical configurations. In this context, the goal of the present work is to validate the pseudo-scattering method against realistic configurations. In this perspective, pebble-bed cores are particularly well-adapted cases to model, as they exhibit large amount of volumes stochastically arranged on two different levels (the pebbles in the core and the TRISO particles inside each pebble). This paper will introduce the techniques and methods used to model pebble-bed cores in a realistic way. The results of the criticality calculations, as well as the pseudo-scattering tracking method performance in terms of computation time, will also be presented. (authors)

  12. Criticality calculations on realistic modelling of pebble-bed HTR-PROTEUS as a validation for the woodcock tracking method implemented in the MORET 5 Monte Carlo code

    International Nuclear Information System (INIS)

    The MORET code is a three dimensional Monte Carlo criticality code. It is designed to calculate the effective multiplication factor (keff) of any geometrical configuration as well as the reaction rates in the various volumes and the neutron leakage out of the system. A recent development for the MORET code consists of the implementation of an alternate neutron tracking method known as the pseudo-scattering tracking method. This method has been successfully implemented in the MORET code and its performances have been tested by the means of an extensive parametric study on very simple geometrical configurations. In this context, the goal of the present work is to validate the pseudo-scattering method against realistic configurations. In this perspective, pebble-bed cores are particularly well-adapted cases to model as they exhibit large amount of volumes stochastically arranged on two different levels (the pebbles in the core and the TRISO particles inside each pebble). This paper will introduce the techniques and methods used to model pebble-bed cores in a realistic way. The results of the criticality calculations, as well as the pseudo-scattering tracking method performance in terms of computation time will be presented. (authors)

  13. Low temperature tritium release experiment from lithium titanate breeder material

    International Nuclear Information System (INIS)

    Engineering data of neutron irradiation performance are needed to design a fusion blanket. Of the engineering data, tritium release characteristic is one of the most important data. Therefore, tritium release experiments of the tritium breeding materials were carried out to evaluate the effects of various parameters, i.e. sweep-gas flow rate, irradiation temperature, hydrogen content in sweep gas and so on, on tritium release. Lithium titanate (Li2TiO3) is a candidate tritium breeding material for the blanket design of International Thermonuclear Experimental Reactor (ITER). As for the shape of the breeder material, a small spherical form is preferred to enhance tritium release from the breeder and to reduce the induced thermal stress in the breeder. Li2TiO3 pebbles with a diameter of 1mm and a total weight of ∼134g have been fabricated, and a pebble-pac assembly of the Li2TiO3 pebbles was irradiated in the Japan Materials Testing Reactor (JMTR), for 3 cycles (about 75 days). The tritium generated in breeder, and released from the breeder was swept downstream by the sweep gas for on-line analysis of tritium content. The total concentration and gaseous concentration of tritium released from the Li2TiO3 pebbles were measured, and HT/(HT+HTO) ratio was evaluated. The sweep-gas flow rate was changed from 10 to 1,000cm3/min, and hydrogen concentration in the sweep gas was changed from 100 to 10,000 ppm. The irradiation temperature of the outer region of the pebble-pac assembly was held below 450degC. The results showed that tritium release from the Li2TiO3 pebbles was started between 100 and 140degC and that the amount of released with increasing the irradiation temperature. The sweep-gas flow rate did not have an effect on tritium release from the Li2TiO3 pebble bed in the steady state. On the other hand, the hydrogen content in the sweep gas had an effect on the tritium release from the Li2TiO3 pebble bed. (author)

  14. 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)

    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

  15. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 3. Appendix A. Equipment list

    International Nuclear Information System (INIS)

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system and was prepared by the General Electric Company. Core scoping studies were performed which evaluated the effects of annular and cylindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations. Volume 3 is an Appendix containing the equipment list for the plant and was also prepared by United Engineers and Constructors, Inc. It tabulates the major components of the plant and describes each in terms of quantity, type, orientation, etc., to provide a basis for cost estimation

  16. Characteristic behavior of pebble-bed modular high-temperature gas-cooled reactor during loss of forced cooling accidents

    International Nuclear Information System (INIS)

    Based on the preliminary design of the Pebble-bed Modular High-Temperature Gas-cooled Reactor(HTR-PM), two cases of loss of forced cooling accident (DLOFC and PLOFC) were studied by the help of the software THERMIX. The key parameters including reactor power, temperature distributions of the core and pressure vessel, and the decay power removal by the passive residual heat remove system(RHRS) were compared in detail. Some parameter uncertainties were analyzed in order to evaluate the safety margin of the maximal fuel temperature during LOFC. The calculated results show that, the decay heat in the LOFC accidents can be removed from the reactor core solely by means of physical processes in a passive way, so that the temperature limits of fuel and components are still obeyed, which can effectively keep the integrality of the fuel particles to avoid massive fission products release. It also illustrates that the HTR-PM can reach 250 MW reactor power per unit and still can keep the inherent safety, which will be helpful to the further detail design of the HTR-PM demonstrating power plant project. (authors)

  17. A comparison of experimental thermal stratification parameters for an oil/pebble-bed thermal energy storage (TES) system during charging

    International Nuclear Information System (INIS)

    Highlights: → Six experimental thermal stratification parameters are evaluated in a TES system. → Stratification number and temperature difference evaluate stratification adequately. → Exergy efficiency and Reynolds number evaluate stratification qualitatively. → Richardson number and energy efficiency not clearly related with stratification. -- Abstract: Six different experimental thermal stratification evaluation parameters during charging for an oil/pebble-bed TES system are presented. The six parameters are the temperature distribution along the height of the storage tank at different time intervals, the charging energy efficiency, the charging exergy efficiency, the stratification number, the Reynolds number and the Richardson number. These parameters are evaluated under six different experimental charging conditions. Temperature distribution along the height of the storage tank at different time intervals and the stratification number are parameters found to describe thermal stratification quantitatively adequately. On the other-hand, the charging exergy efficiency and the Reynolds number give important information about describing thermal stratification qualitatively and should be used with care. The charging energy efficiency and the Richardson number have no clear relationship with thermal stratification.

  18. 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)

    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

  19. Proceedings of the eleventh international workshop on ceramic breeder blanket interactions

    International Nuclear Information System (INIS)

    This report is the Proceedings of 'the Eleventh International Workshop on Ceramic Breeder Blanket Interactions' which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors, and the Japan-US Fusion Collaboration Framework. This workshop was held in Tokyo, Japan on December 15-17, 2003. About thirty experts from China, EU, Japan, Korea, Latvia, Russia and USA attended the workshop. The scope of the workshop included 1) evolutions in ceramic breeder blanket design, 2) progress in ceramic breeder material development, 3) irradiation testing, 4) breeder material properties, 5) out-of-pile pebble bed experiment, 6) modeling of the thermal, mechanical and tritium transfer behavior of pebble beds and 7) interfacing issues of solid breeder blanket. In the workshop, information exchange was performed for designs of solid breeder blankets and test blankets in EU, Russia and Japan, recent results of irradiation tests, HICU, EXOTIC-8 and the irradiation tests by IVV-2M, modeling study on tritium release behavior of Li2TiO3 and so on, fabrication technology developments and characterization of the Li2TiO3 and Li4SiO4 pebbles, research on measurements and modeling of thermo-mechanical behaviors of Li2TiO3 and Li4SiO4 pebbles, and interfacing issues, such as, fabrication technology for blanket box structure, neutronics experiments of blanket mockups by fusion neutron source and tritium recovery system. The 26 of the presented papers are indexed individually. (J.P.N.)

  20. A generalized theory for non-classical transport with angular-dependent path-length distributions 2: Anisotropic diffusion in model pebble bed reactor cores

    CERN Document Server

    Vasques, Richard

    2013-01-01

    We describe an analysis of neutron transport in the interior of model pebble bed reactor (PBR) cores, considering both crystal and random pebble arrangements. Monte Carlo codes were developed for (i) generating random realizations of the model PBR core, and (ii) performing neutron transport inside the crystal and random heterogeneous cores; numerical results are presented for two different choices of material parameters. These numerical results are used to investigate the anisotropic behavior of neutrons in each case and to assess the accuracy of estimates for the diffusion coefficients obtained with the diffusion approximations of different models: the atomic mix model, the Behrens correction, the Lieberoth correction, the generalized linear Boltzmann equation (GLBE), and the new GLBE with angular-dependent path-length distributions. This new theory utilizes a non-classical form of the Boltzmann equation in which the locations of the scattering centers in the system are correlated and the distance-to-collisi...

  1. Fission product transport in the primary system of a pebble bed high temperature reactor with direct cycle

    International Nuclear Information System (INIS)

    Transport and deposition of fission products in the primary system of a small pebble bed high temperature reactor with directly coupled gas turbine have been investigated. The reactor has a thermal power of 40 MW and is intended for heat and power cogeneration. Four radionuclides have been identified as most relevant because of volatility and radiotoxicity: 137Cs, 90Sr, 110mAg, 131I. With the code PANAMA the fraction of failed coated fuel particles has been calculated. The diffusion of the fission products to the fuel element outer surface has been calculated with the FRESCO code. Transport and deposition of the fission products within the primary system has been analysed with the code MELCOR. Under normal operating conditions the release rate of the short-lived 131I reaches a constant level rather quickly, contrary to the longer lived 137Cs and 90Sr which show a steady increase of the release rate during burn-up. Under incident conditions the retention capability of the fuel elements' graphite is strongly reduced. The release from the intact coated particles remains negligible compared to the release from the defect coated particles. After a ten year operation period, the total activity of the released nuclides in the primary system is about 58 GBq. The highest activity is found in the pre-cooler. Other components with high activities are the recuperator and the compressor. These components are contaminated mainly by 110mAg. The gas ducts in the energy conversion unit are contaminated by 110mAg and 131I. Contamination as a consequence of incident conditions is difficult to estimate, because it depends on a number of phenomena. Under the assumption that 10 fuel elements are damaged, the activity is about 44 GBq. (author)

  2. 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)

    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 keff 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 keff on fuel kernel radius decreases in relation to the increase in uranium enrichment . However, keff 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)

  3. The analytic function expansion nodal (AFEN) method with half-interface averaged fluxes in mixed geometry nodes for analysis of pebble bed modular reactor (PBMR) cores

    International Nuclear Information System (INIS)

    The analytic function expansion nodal (AFEN) method has been successfully applied to the rectangular and hexagonal geometries in the cartesian coordinates system. In this paper, we extended the AFEN method to the cylindrical geometry in the R-Z coordinates for the analysis of pebble bed modular reactors (PBMRs). To treat the mixed geometry of rectangular and triangular nodes appearing in the lower periphery of the reactors, we used half-interface averaged fluxes as nodal unknowns. Numerical results obtained attest to their accuracy and applicability to practical problems. (author)

  4. Disposition of plutonium with HTGRs using Pu burner balls and Th breeder balls

    International Nuclear Information System (INIS)

    A concept of reactor system was developed with which weapons-grade Plutonium could be made perfectly worthless in use for weapons. It is a pebble bed type HTGR using Pu burner ball fuels and Th breeder ball fuels. The residual amounts of 239Pu in spent Pu balls become less than 1% of the initial loading. The power coefficient was made negative by reducing the parasitic neutron absorption reaction rate of 135Xe. (author)

  5. Modeling of the Fluid Flow and Heat Transfer an a Pebble Bed Modular Reactor Core With a Computational Fluid Dynamics Code

    International Nuclear Information System (INIS)

    The Pebble Bed Modular Reactor (PBMR), a promising Generation IV nuclear reactor design, raises many novel technological issues for which new experience and techniques must be developed. This brief study explores a few of these issues, utilizes a computational fluid dynamics code to model some simple phenomena, and points out deficiencies in current knowledge that should be addressed by future research and experimentation. A highly simplified representation of the PBMR core is analyzed with FLUENT, a commercial computational fluid dynamics code. The applied models examine laminar and turbulent flow in the vicinity of a single spherical fuel pebble near the center of the core, accounting for the effects of the immediately adjacent fuel pebbles. Several important fluid flow and heat transfer parameters are examined, including heat transfer coefficient, Nusselt number, and pressure drop, as well as the temperature, pressure, and velocity profiles near the fuel pebble. The results of these 'unit cell' calculations are also compared to empirical correlations available in the literature. As FLUENT is especially sensitive to geometry during the generation of a computational mesh, the sensitivity of code results to pebble spacing is also examined. The results of this study show that while a PBMR presents a novel and complex geometry, a code such as FLUENT is suitable for calculation of both local and global flow characteristics, and can be a valuable tool for the thermal-hydraulic study of this new reactor design. FLUENT results for pressure drop deviate from the Darcy correlation by several orders of magnitude in all cases. When determining the heat transfer coefficient, FLUENT is again much lower than Robinson's correlation. Results for Nusselt number show better agreement, with FLUENT predicting results that are 10 or 20 times as large as those from the Robinson and Lancashire correlations. These differences may arise because the empirical correlations concern mainly

  6. Preliminary structural design and thermo-mechanical analysis of helium cooled solid breeder blanket for Chinese Fusion Engineering Test Reactor

    International Nuclear Information System (INIS)

    Highlights: • A helium cooled solid breeder blanket module was designed for CFETR. • Multilayer U-shaped pebble beds were adopted in the blanket module. • Thermal and thermo-mechanical analyses were carried out under normal operating conditions. • The analysis results were found to be acceptable. - Abstract: With the aim to bridge the R&D gap between ITER and fusion power plant, the Chinese Fusion Engineering Test Reactor (CFETR) was proposed to be built in China. The mission of CFETR is to address the essential R&D issues for achieving practical fusion energy. Its blanket is required to be tritium self-sufficient. In this paper, a helium cooled solid breeder blanket adopting multilayer U-shaped pebble beds was designed and analyzed. Thermo-mechanical analysis of the first wall and side wall combined with breeder unit was carried out for normal operating steady state conditions. The results showed that the maximum temperatures of the structural material, neutron multiplier and tritium breeder pebble beds are 523 °C, 558 °C and 787 °C, respectively, which are below the corresponding limits of 550 °C, 650 °C and 920 °C. The maximum equivalent stress of the structure is under the allowable value with a margin about 14.5%

  7. Design and construction of a prototype advanced on-line fuel burn-up monitoring system for the modular pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Su, Bingjing; Hawari, Ayman, I.

    2004-03-30

    Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor currently under study as a next generation reactor system. In addition to its inherently safe design, a unique feature of this reactor is its multi-pass fuel circulation in which the fuel pebbles are randomly loaded and continuously cycled through the core until they reach their prescribed End-of-Life burn-up limit. Unlike the situation with a conventional light water reactor, depending solely on computational methods to perform in-core fuel management for MPBR will be highly inaccurate. An on-line measurement system is needed to accurately assess whether a given pebble has reached its End-of-Life burn-up limit and thereby provide an on-line, automated go/no-go decision on fuel disposition on a pebble-by-pebble basis. This project investigated approaches to analyzing fuel pebbles in real time using gamma spectroscopy and possibly using passive neutron counting of spontaneous fission neutrons to provide the speed, accuracy, and burn-up range required for burnup determination of MPBR. It involved all phases necessary to develop and construct a burn-up monitor, including a review of the design requirements of the system, identification of detection methodologies, modeling and development of potential designs, and finally, the construction and testing of an operational detector system. Based upon the research work performed in this project, the following conclusions are made. In terms of using gamma spectrometry, two possible approaches were identified for burnup assay. The first approach is based on the measurement of the absolute activity of Cs-137. However, due to spectral interference and the need for absolute calibration of the spectrometer, the uncertainty in burnup determination using this approach was found to range from {approx} {+-}40% at beginning of life to {approx} {+-}10% at the discharge burnup. An alternative approach is to use a relative burnup indicator. In this

  8. Design and construction of a prototype advanced on-line fuel burn-up monitoring system for the modular pebble bed reactor

    International Nuclear Information System (INIS)

    Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor currently under study as a next generation reactor system. In addition to its inherently safe design, a unique feature of this reactor is its multi-pass fuel circulation in which the fuel pebbles are randomly loaded and continuously cycled through the core until they reach their prescribed End-of-Life burn-up limit. Unlike the situation with a conventional light water reactor, depending solely on computational methods to perform in-core fuel management for MPBR will be highly inaccurate. An on-line measurement system is needed to accurately assess whether a given pebble has reached its End-of-Life burn-up limit and thereby provide an on-line, automated go/no-go decision on fuel disposition on a pebble-by-pebble basis. This project investigated approaches to analyzing fuel pebbles in real time using gamma spectroscopy and possibly using passive neutron counting of spontaneous fission neutrons to provide the speed, accuracy, and burn-up range required for burnup determination of MPBR. It involved all phases necessary to develop and construct a burn-up monitor, including a review of the design requirements of the system, identification of detection methodologies, modeling and development of potential designs, and finally, the construction and testing of an operational detector system. Based upon the research work performed in this project, the following conclusions are made. In terms of using gamma spectrometry, two possible approaches were identified for burnup assay. The first approach is based on the measurement of the absolute activity of Cs-137. However, due to spectral interference and the need for absolute calibration of the spectrometer, the uncertainty in burnup determination using this approach was found to range from ∼ ±40% at beginning of life to ∼ ±10% at the discharge burnup. An alternative approach is to use a relative burnup indicator. In this case, a self

  9. Proposed chemical plant initiated accident scenarios in a sulphur-iodine cycle plant coupled to a pebble bed modular reactor

    International Nuclear Information System (INIS)

    In the sulphur-iodine (S-I) cycle nuclear hydrogen generation scheme the chemical plant acts as the heat sink for the very high temperature nuclear reactor (VHTR). Thus, any accident which occurs in the chemical plant must feedback to the nuclear reactor. There are many different types of accidents which can occur in a chemical plant. These accidents include intra-reactor piping failure, inter-reactor piping failure, reaction chamber failure and heat exchanger failure. Since the chemical plant acts as the heat sink for the nuclear reactor, any of these accidents induce a loss-of-heat-sink accident in the nuclear reactor. In this paper, several chemical plant initiated accident scenarios are presented. The following accident scenarios are proposed: i) failure of the Bunsen chemical reactor; ii) product flow failure from either the H2SO4 decomposition section or HI decomposition section; iii) reactant flow failure from either the H2SO4 decomposition section or HI decomposition section; iv) rupture of a reaction chamber. Qualitative analysis of these accident scenarios indicates that each result in either partial or total loss of heat sink accidents for the nuclear reactor. These scenarios are reduced to two types: i) discharge rate limited accidents; ii) discontinuous reaction chamber accidents. A discharge rate limited rupture of the SO3 decomposition section of the SI cycle is proposed and modelled. Since SO3 decomposition occurs in the gaseous phase, critical flow out of the rupture is calculated assuming ideal gas behaviour. The accident scenario is modelled using a fully transient control volume model of the S-I cycle coupled to a THERMIX model of a 268 MW pebble bed modular reactor (PBMR-268) and a point kinetics model. The Bird, Stewart and Lightfoot source model for choked gas flows from a pressurised chamber was utilised as a discharge rate model. A discharge coefficient of 0.62 was assumed. Feedback due to the rupture is observed in the nuclear reactor

  10. SHOVAV-JUEL. A one dimensional space-time kinetic code for pebble-bed high-temperature reactors with temperature and Xenon feedback

    International Nuclear Information System (INIS)

    The present report describes the modelling basis and the structure of the neutron kinetics-code SHOVAV-Juel. Information for users is given regarding the application of the code and the generation of the input data. SHOVAV-Juel is a one-dimensional space-time-code based on a multigroup diffusion approach for four energy groups and six groups of delayed neutrons. It has been developed for the analysis of the transient behaviour of high temperature reactors with pebble-bed core. The reactor core is modelled by horizontal segments to which different materials compositions can be assigned. The temperature dependence of the reactivity is taken into account by using temperature dependent neutron cross sections. For the simulation of transients in an extended time range the time dependence of the reactivity absorption by Xenon-135 is taken into account. (orig./RW)

  11. User's manual for ASTERIX-2: A two-dimensional modular code system for the steady state and xenon transient analysis of a pebble bed high temperature reactor

    International Nuclear Information System (INIS)

    The ASTERIX modular code package was developed at KFA Laboratory-Juelich for the steady state and xenon transient analysis of a pebble bed high temperature reactor. The code package was implemented on the Stanford Linear Accelerator Center Computer in August, 1980, and a user's manual for the current version of the code, identified as ASTERIX-2, was prepared as a cooperative effort by KFA Laboratory and GE-ARSD. The material in the manual includes the requirements for accessing the program, a description of the major subroutines, a listing of the input options, and a listing of the input data for a sample problem. The material is provided in sufficient detail for the user to carry out a wide range of analysis from steady state operations to the xenon induced power transients in which the local xenon, temperature, buckling and control feedback effects have been incorporated in the problem solution. (orig.)

  12. Analytical Solution of Fick's Law of the TRISO-Coated Fuel Particles and Fuel Elements in Pebble-Bed High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Two kinds of approaches are built to solve the fission products diffusion models (Fick's equation) based on sphere fuel particles and sphere fuel elements exactly. Two models for homogenous TRISO-coated fuel particles and fuel elements used in pebble-bed high temperature gas-cooled reactors are presented, respectively. The analytical solution of Fick's equation for fission products diffusion in fuel particles is derived by variables separation. In the fuel element system, a modification of the diffusion coefficient from D to D/r is made to characterize the difference of diffusion rates in distinct areas and it is shown that the Laplace and Hankel transformations are effective as the diffusion coefficient in Fick's equation is dependant on the radius of the fuel element. Both the solutions are useful for the prediction of the fission product behaviors and could be programmed in the corresponding engineering calculations. (general)

  13. Analytical solution of Fick's law of the TRISO-coated fuel particles and fuel elements in pebble-bed high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Two kinds of approaches are built to solve the fission products diffusion models (Fick's equation) based on sphere fuel particles and sphere fuel elements exactly. Two models for homogenous TRISO-coated fuel particles and fuel elements used in pebble-bed high temperature gas-cooled reactors are presented, respectively. The analytical solution of Fick's equation for fission products diffusion in fuel particles is derived by variables separation. In the fuel element system, a modification of the diffusion coefficient from D to D/r is made to characterize the difference of diffusion rates in distinct areas and it is shown that the Laplace and Hankel transformations are effective as the diffusion coefficient in Fick's equation is dependant on the radius of the fuel element. Both the solutions are useful for the prediction of the fission product behaviors and could be programmed in the corresponding engineering calculations. (authors)

  14. Analytical Solution of Fick's Law of the TRISO-Coated Fuel Particles and Fuel Elements in Pebble-Bed High Temperature Gas-Cooled Reactors

    Institute of Scientific and Technical Information of China (English)

    CAO Jian-Zhu; FANG Chao; SUN Li-Feng

    2011-01-01

    T wo kinds of approaches are built to solve the fission products diffusion models (Fick's equation) based on sphere fuel particles and sphere fuel elements exactly. Two models for homogenous TRISO-coated fuel particles and fuel elements used in pebble-bed high temperature gas-cooled reactors are presented, respectively. The analytica,solution of Fick's equation for fission products diffusion in fuel particles is derived by variables separation.In the fuel element system, a modification of the diffusion coefficient from D to D/r is made to characterize the difference of diffusion rates in distinct areas and it is shown that the Laplace and Hankel transformations are effective as the diffusion coefficient in Fick's equation is dependant on the radius of the fuel element. Both the solutions are useful for the prediction of the fission product behaviors and could be programmed in the corresponding engineering calculations.%@@ Two kinds of approaches are built to solve the fission products diffusion models(Fick's equation) based on sphere fuel particles and sphere fuel elements exactly.Two models for homogenous TRISO-coated fuel particles and fuel elements used in pebble-bed high temperature gas-cooled reactors are presented,respectively.The analytical solution of Fick's equation for fission products diffusion in fuel particles is derived by variables separation.In the fuel element system,a modification of the diffusion coefficient from D to D/r is made to characterize the difference of diffusion rates in distinct areas and it is shown that the Laplace and Hankel transformations are effective as the diffusion coefficient in Fick's equation is dependant on the radius of the fuel element.Both the solutions are useful for the prediction of the fission product behaviors and could be programmed in the corresponding engineering calculations.

  15. Construction of PREMUX and preliminary experimental results, as preparation for the HCPB breeder unit mock-up testing

    International Nuclear Information System (INIS)

    Highlights: • PREMUX has been constructed as preparation for a future out-of-pile thermo-mechanical qualification of a HCPB breeder unit mock-up. • The rationale and constructive details of PREMUX are reported in this paper. • PREMUX serves as a test rig for the new heater system developed for the HCPB-BU mock-up. • PREMUX will be used as benchmark for the thermal and thermo-mechanical models developed in ANSYS for the pebble beds of the HCPB-BU. • Preliminary results show the functionality of PREMUX and the good agreement of the measured temperatures with the thermal model developed in ANSYS. - Abstract: One of the European blanket designs for ITER is the Helium Cooled Pebble Bed (HCPB) blanket. The core of the HCPB-TBM consists of so-called breeder units (BUs), which encloses beryllium as neutron multiplier and lithium orthosilicate (Li4SiO4) as tritium breeder in form of pebble beds. After the design phase of the HCPB-BU, a non-nuclear thermal and thermo-mechanical qualification program for this device is running at the Karlsruhe Institute of Technology. Before the complex full scale BU testing, a pre-test mock-up experiment (PREMUX) has been constructed, which consists of a slice of the BU containing the Li4SiO4 pebble bed. PREMUX is going to be operated under highly ITER-relevant conditions and has the following goals: (1) as a testing rig of new heater concept based on a matrix of wire heaters, (2) as benchmark for the existing finite element method (FEM) codes used for the thermo-mechanical assessment of the Li4SiO4 pebble bed, and (3) in situ measurement of thermal conductivity of the Li4SiO4 pebble bed during the tests. This paper describes the construction of PREMUX, its rationale and the experimental campaign planned with the device. Preliminary results testing the algorithm used for the temperature reconstruction of the pebble bed are reported and compared qualitatively with first analyses completed with the FEM codes

  16. Influence of power level and fuel type on safety and economy of the simplified pebble bed HTR concept

    International Nuclear Information System (INIS)

    For three different power levels, 20, 40 and 150 MWth, the PAP-HTR has been studied. This is an HTR Module concept that has been simplified in such a way that the continuously defuelling system has been eliminated and no defuelling takes place during a period of several years. Two core heatup scenarios have been simulated. It has been shown that in all cases the maximum fuel pebble temperature remains below 1600C, the temperature above which fuel degradation would start to occur, also after the reactor has gone critical again and the power level has been stabilized by itself. Fuel and gas temperature distributions are compared as well. The maximum pebble temperature before recriticality is higher for the loss of coolant (LOCA) scenario than for the loss of flow (LOFA) case, but the equilibrium maximum temperature after recriticality turns out to be higher for the pressurized case, because of the higher equilibrium power level. The equilibrium power level is a much smaller fraction of the nominal power level for the large 150 MWth system than for the smaller systems, due to the lower rate of cooling down of the large system after initiation of the accident. Therefore the equilibrium maximum temperature stays within acceptable limits for the large system too. The effects of the use of thorium fuel on the core height and waste radiotoxicity have been compared with the case of uranium fuel. Although it is widely believed that burnt thorium fuel would be cleaner than spent uranium fuel in terms of radiotoxicity, this did not appear to be more pronounced for this reactor concept than for e.g. PWRs. The relationship of power level and energy price is obvious for this power range. The use of thorium with highly enriched uranium could bring an additional economical advantage because of the lower core height needed for the same power level as the uranium case. With thorium a higher burnup can be attained, through which fuel pebbles can be added at a slower rate. The size of

  17. Large modular pebble-bed reactors with passive safety properties as a contribution for catastrophe-free nuclear technology. Flexibility in design and application

    International Nuclear Information System (INIS)

    Worldwide investigations are carried out for different reactor concepts, in order to realize nuclear energy production in modular power plants. In that concept several small or middle sized reactors are joined together in a modular way to form one power plant. The size of MODUL-reactors is designed in such a way, that exclusively inherent safety properties perform the control of accidents without active technical proceedings. In order to achieve this, the reactor should be relatively small. On the other hand, it should be relatively large for economic and competitive reasons. The range of possible development of the modular pebble-bed reactor for raising the power output are discussed in this study. Based on the MODUL 200 MW concept, the design of the 'Great-Modul-Medul' reactor (GMM) with a power output of 500 MWth is introduced, in which the loading modus MEDUL is applied with repeated circulation of the spheres through the core. A 'Great-Modul-OTTO' GMO with a power output of 400 MWth is designed with only one pass of the pebbles (OTTO). In comparison to the GMM, that has the advantage of being simpler in construction and in the method of operation. Furthermore, another simplification is studied consisting of the combination (PO) of 'Peu a Peu' and 'OTTO' loading modus. All designed cases show a favourable flexibility when changing the application of the reactor from steam cycle to gas turbine cycle or to seawater desalination. The study outlines, that the inherently determined limitation of the excess temperature in case of a loss coolant accident and the ability for controling the water ingress reactivity are maintained for all variants being considered. (orig.)

  18. 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)

    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.

  19. Radiolysis of Slightly Overstoichiometric Lithium Orthosilicate Pebbles

    OpenAIRE

    Zarins, A.; Supe, A; Kizane, G; Knitter, R.; Reinholds, I; Vitins, A; Tilika, V; Actins, A; Baumane, L

    2010-01-01

    : One of the technological problems of a fusion reactor is the change in composition and structure of ceramic breeder (Li4SiO4 or Li2TiO3 pebbles) during long-term operation. Changes in the composition and structure of the Li4SiO4 ceramic pebbles at fast electron irradiation (E = 5 MeV, dose rate up to 88 MGy•h-1, absorbed dose up to 10.6 GGy) at 543-573 K were investigated in this study. Overstoichiometric (2.5 weight % of additional SiO2) lithium orthosilicate pebbles were fabricated by...

  20. 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.

  1. 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.

  2. New progress on design and R and D for solid breeder test blanket module in China

    Energy Technology Data Exchange (ETDEWEB)

    Feng, K.M., E-mail: fengkm@swip.ac.cn; Zhang, G.S.; Hu, G.; Chen, Y.J.; Feng, Y.J.; Li, Z.X.; Wang, P.H.; Zhao, Z.; Ye, X.F.; Xiang, B.; Zhang, L.; Wang, Q.J.; Cao, Q.X.; Zhao, F.C.; Wang, F.; Liu, Y.; Zhang, M.C.

    2014-10-15

    Highlights: • The new progress on design and R and D of Chinese solid breeder TBM are introduced. • The mock-up fabrication and component tests for Chinese HCCB TBM have being developed. • The neutron multiplier Be pebbles, tritium breeder Li{sub 4}SiO{sub 4} pebbles, and structure material CFL-1 are being prepared. • The fabrication of 1/3 sized mock-up is being carried-out. • The key technology development is proceeding to the large-scale mock-up fabrication. - Abstract: ITER will be used to test tritium breeding module concepts, which will lead to the design of DEMO fusion reactor demonstrating tritium self-sufficiency and the extraction of high grade heat for electricity production. China plans to test the HCCB TBM modules during different operation phases. Related design and R and D activities for each TBM module with the auxiliary system are introduced. The helium-cooled ceramic breeder (HCCB) test blanket module (TBM) is the primary option of the Chinese TBM program. The preliminary conceptual design of CN HCCB TBM has been completed. A modified design to reduce the RAFM material mass to 1.3 ton has been carried out based on the ITER technical requirement. Basic characteristics and main design parameters of CN HCCB TBM are introduced briefly. The mock-up fabrication and component tests for Chinese test blanket module are being developed. Recent status of the components of CN HCCB TBM and fabrication technology development are also reported. The neutron multiplier Be pebbles, tritium breeder Li{sub 4}SiO{sub 4} pebbles, and structure material CLF-1 of ton-class are being prepared in laboratory scale. The fabrication of pebble bed container and experiment of tritium breeder pebble bed will be started soon. The fabrication technology development is proceeding as the large-scale mock-up fabrication enters into the R and D stage and demonstration tests toward TBM testing on ITER test port are being done as scheduled.

  3. The Karlsruhe solid breeder blanket and the test module to be irradiated in ITER/NET

    International Nuclear Information System (INIS)

    The blanket for the DEMO reactor should operate at an average neutron flux of 2.2 MW/m2 for 20000 h. This requires the use of a structural material which can withstand high neutron fluences without swelling. The ferritic steel Manet was chosen for this purpose. The breeder material is in the form of Li4SiO4 pebbles of 0.35 to 0.6 mm diameter. The 6 mm thick beds of pebbles are placed between beryllium plates which are cooled by high pressure helium flowing inside steel tubes. Breeder material and beryllium are contained in radial canisters, placed inside boxes. The coolant helium enters the blanket at 250deg C, cools first the box walls and then the breeder and multiplier, and leaves the blanket at 450deg C. The maximum temperature in the first wall steel is 550deg C, while the minimum and maximum temperatures in the breeder are 380 and 820deg C, respectively. The resulting total tritium inventory in the breeder is only 10 g, and the real tridimensional tritium breeding ratio is 1.11. The conceptual design of the test module, of its extraction system and of the required out-of-reactor ancillary systems has allowed an estimate of the time constants of the various components and thus allowed an assessment of the requirements given by the testing of the modules on the NET/ITER machine. (orig.)

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

    International Nuclear Information System (INIS)

    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.)

  5. 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)

    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.)

  6. Feasibility study of a fission-suppressed tokamak fusion breeder

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.; Lee, J.D.; Neef, W.S.; Berwald, D.H.; Garner, J.K.; Whitley, R.H.; Ghoniem, N.; Wong, C.P.C.; Maya, I.; Schultz, K.R.

    1984-12-01

    The preliminary conceptual design of a tokamak fissile fuel producer is described. The blanket technology is based on the fission suppressed breeding concept where neutron multiplication occurs in a bed of 2 cm diameter beryllium pebbles which are cooled by helium at 50 atmospheres pressure. Uranium-233 is bred in thorium metal fuel elements which are in the form of snap rings attached to each beryllium pebble. Tritium is bred in lithium bearing material contained in tubes immersed in the pebble bed and is recovered by a purge flow of helium. The neutron wall load is 3 MW/m/sup 2/ and the blanket material is ferritic steel. The net fissile breeding ratio is 0.54 +- 30% per fusion reaction. This results in the production of 4900 kg of /sup 233/U per year from 3000 MW of fusion power. This quantity of fuel will provide makeup fuel for about 12 LWRs of equal thermal power or about 18 1 GW/sub e/ LWRs. The calculated cost of the produced uranium-233 is between $23/g and $53/g or equivalent to $10/kg to $90/kg of U/sub 3/O/sub 8/ depending on government financing or utility financing assumptions. Additional topics discussed in the report include the tokamak operating mode (both steady state and long pulse considered), the design and breeding implications of using a poloidal divertor for impurity control, reactor safety, the choice of a tritium breeder, and fuel management.

  7. Feasibility study of a fission-suppressed tokamak fusion breeder

    International Nuclear Information System (INIS)

    The preliminary conceptual design of a tokamak fissile fuel producer is described. The blanket technology is based on the fission suppressed breeding concept where neutron multiplication occurs in a bed of 2 cm diameter beryllium pebbles which are cooled by helium at 50 atmospheres pressure. Uranium-233 is bred in thorium metal fuel elements which are in the form of snap rings attached to each beryllium pebble. Tritium is bred in lithium bearing material contained in tubes immersed in the pebble bed and is recovered by a purge flow of helium. The neutron wall load is 3 MW/m2 and the blanket material is ferritic steel. The net fissile breeding ratio is 0.54 +- 30% per fusion reaction. This results in the production of 4900 kg of 233U per year from 3000 MW of fusion power. This quantity of fuel will provide makeup fuel for about 12 LWRs of equal thermal power or about 18 1 GW/sub e/ LWRs. The calculated cost of the produced uranium-233 is between $23/g and $53/g or equivalent to $10/kg to $90/kg of U3O8 depending on government financing or utility financing assumptions. Additional topics discussed in the report include the tokamak operating mode (both steady state and long pulse considered), the design and breeding implications of using a poloidal divertor for impurity control, reactor safety, the choice of a tritium breeder, and fuel management

  8. Li4SiO4 pebbles reduction in He + 0.1% H2 purge gas and effects on tritium release properties

    International Nuclear Information System (INIS)

    Lithium orthosilicate reduction was examined by Temperature Programmed Reaction (TPR) and Temperature Programmed Desorption (TPD) methods performed in He (or Ar) + H2 purge gas flowing through pebble bed specimens. The parameters governing the kinetics and the steady-state of the reduction process to Li4SiO4-x were determined at 800 deg. C. The level x of the O-vacancy concentration at steady-state (of the order of 1.5x10-3 mole fraction) was found to be compatible with the impurities content in the specimens. Pebble pre-annealing treatments were found to affect the microstructure and the reduction mechanism. Post-irradiation tritium release by TPD tests were performed on both stoichiometric and reduced pebbles with similar results. Tritium release properties of this breeder system seem to be independent from the material reduction state (x)

  9. 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

    2015-01-01

    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.

  10. Experimental investigations of graphite corrosion and aerosol formation during air ingress into the core of a high temperature pebble bed reactor

    International Nuclear Information System (INIS)

    A High Temperature Reactor can be designed to remove the decay heat without using any active systems. For most accident scenarios a release of radioactive fission products can be excluded by design. However, during operation of a HTR some accidents are principally possible, which can result in a release of fission products out of the fuel elements and of the reactor system. One of these accidents is a hypothetical massive air ingress into the hot graphite reactor core. After a pressure drop caused by leakages in the primary circuit a gas mass flow may be able to stream through the core according to free natural convection leading to a corrosion of the graphite fuel elements and reflector structures. With the VELUNA-experiment a testing device was installed, which allows to investigate the corrosion process on parts of a reactor core under real accident conditions. With regard to the experimental results already existing equations to calculate the chemical reaction rate in a pebble bed were modified and the applicability was demonstrated. These equations consider the chemical reaction in the porous graphite as well as diffusion processes to the graphite surface. Equivalent correlations were developed for different flow geometries and for the graphite material of the bottom reflector. The corrosion process forms an aerosol, which consists of graphite particles and a reaction gas phase. The formatted aerosol was characterized concerning its chemical and physical properties. Because the aerosol particles can support the release of fission products, measurements of aerosol parameters like particle mass concentration and particle size distribution provide important information to estimate the radiologic consequences of such an hypothetical air ingress accident. (orig.)

  11. Experimental and numerical investigations of heat transfer in the first wall of Helium-Cooled-Pebble-Bed Test Blanket Module—Part 2: Presentation of results

    International Nuclear Information System (INIS)

    Highlight: • Results of heat transfer investigations in the first wall are presented. - Abstract: This paper is the continuation of the first report on investigations of heat transfer in the first wall of Helium-Cooled-Pebble-Bed Test Blanket Module for ITER submitted to this Journal (see Ilić et al. [1]). The investigations have been performed experimentally by manufacturing and testing of a mock-up and numerically through the development of corresponding 3D CFD models. The experimental tests have been conducted for HCPB TBM relevant conditions – the test channel made of Eurofer steel, helium coolant at pressure of 8 MPa and inlet temperature of 300 °C and heat flux of 270 kW/m2 at the channel side representing plasma facing side of the first wall. In total six measuring series have been performed in which surface roughness, helium inlet temperature and heater power have been considered as parameters. For each measuring series corresponding 3D CFD computational scenarios have been conducted. By use of experimental data for Eurofer temperature it was possible to verify 3D CFD models. On the other side, 1D CFD approaches failed in comparison with experimental data. Further, a critical analysis of the use of microscopic surface roughness as a method for heat transfer improvement in the first wall has been presented. Finally, based on a detailed analysis of experimental and 3D CFD data obtained in the framework of these activities the main directions for an improvement of cooling channels in the first wall could be proposed

  12. Development Status of the PEBBLES Code for Pebble Mechanics: Improved Physical Models and Speed-up

    Energy Technology Data Exchange (ETDEWEB)

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2009-12-01

    PEBBLES is a code for simulating the motion of all the pebbles in a pebble bed reactor. Since pebble bed reactors are packed randomly and not precisely placed, the location of the fuel elements in the reactor is not deterministically known. Instead, when determining operating parameters the motion of the pebbles can be simulated and stochastic locations can be found. The PEBBLES code can output information relevant for other simulations of the pebble bed reactors such as the positions of the pebbles in the reactor, packing fraction change in an earthquake, and velocity profiles created by recirculation. The goal for this level three milestone was to speedup the PEBBLES code through implementation on massively parallel computer. Work on this goal has resulted in speeding up both the single processor version and creation of a new parallel version of PEBBLES. Both the single processor version and the parallel running capability of the PEBBLES code have improved since the fiscal year start. The hybrid MPI/OpenMP PEBBLES version was created this year to run on the increasingly common cluster hardware profile that combines nodes with multiple processors that share memory and a cluster of nodes that are networked together. The OpenMP portions use the Open Multi-Processing shared memory parallel processing model to split the task across processors in a single node that shares memory. The Message Passing Interface (MPI) portion uses messages to communicate between different nodes over a network. The following are wall clock speed up for simulating an NGNP-600 sized reactor. The single processor version runs 1.5 times faster compared to the single processor version at the beginning of the fiscal year. This speedup is primarily due to the improved static friction model described in the report. When running on 64 processors, the new MPI/OpenMP hybrid version has a wall clock speed up of 22 times compared to the current single processor version. When using 88 processors, a

  13. Development Status of the PEBBLES Code for Pebble Mechanics: Improved Physical Models and Speed-up

    Energy Technology Data Exchange (ETDEWEB)

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2009-09-01

    PEBBLES is a code for simulating the motion of all the pebbles in a pebble bed reactor. Since pebble bed reactors are packed randomly and not precisely placed, the location of the fuel elements in the reactor is not deterministically known. Instead, when determining operating parameters the motion of the pebbles can be simulated and stochastic locations can be found. The PEBBLES code can output information relevant for other simulations of the pebble bed reactors such as the positions of the pebbles in the reactor, packing fraction change in an earthquake, and velocity profiles created by recirculation. The goal for this level three milestone was to speedup the PEBBLES code through implementation on massively parallel computer. Work on this goal has resulted in speeding up both the single processor version and creation of a new parallel version of PEBBLES. Both the single processor version and the parallel running capability of the PEBBLES code have improved since the fiscal year start. The hybrid MPI/OpenMP PEBBLES version was created this year to run on the increasingly common cluster hardware profile that combines nodes with multiple processors that share memory and a cluster of nodes that are networked together. The OpenMP portions use the Open Multi-Processing shared memory parallel processing model to split the task across processors in a single node that shares memory. The Message Passing Interface (MPI) portion uses messages to communicate between different nodes over a network. The following are wall clock speed up for simulating an NGNP-600 sized reactor. The single processor version runs 1.5 times faster compared to the single processor version at the beginning of the fiscal year. This speedup is primarily due to the improved static friction model described in the report. When running on 64 processors, the new MPI/OpenMP hybrid version has a wall clock speed up of 22 times compared to the current single processor version. When using 88 processors, a

  14. Development Status of the PEBBLES Code for Pebble Mechanics: Improved Physical Models and Speed-up

    International Nuclear Information System (INIS)

    PEBBLES is a code for simulating the motion of all the pebbles in a pebble bed reactor. Since pebble bed reactors are packed randomly and not precisely placed, the location of the fuel elements in the reactor is not deterministically known. Instead, when determining operating parameters the motion of the pebbles can be simulated and stochastic locations can be found. The PEBBLES code can output information relevant for other simulations of the pebble bed reactors such as the positions of the pebbles in the reactor, packing fraction change in an earthquake, and velocity profiles created by recirculation. The goal for this level three milestone was to speedup the PEBBLES code through implementation on massively parallel computer. Work on this goal has resulted in speeding up both the single processor version and creation of a new parallel version of PEBBLES. Both the single processor version and the parallel running capability of the PEBBLES code have improved since the fiscal year start. The hybrid MPI/OpenMP PEBBLES version was created this year to run on the increasingly common cluster hardware profile that combines nodes with multiple processors that share memory and a cluster of nodes that are networked together. The OpenMP portions use the Open Multi-Processing shared memory parallel processing model to split the task across processors in a single node that shares memory. The Message Passing Interface (MPI) portion uses messages to communicate between different nodes over a network. The following are wall clock speed up for simulating an NGNP-600 sized reactor. The single processor version runs 1.5 times faster compared to the single processor version at the beginning of the fiscal year. This speedup is primarily due to the improved static friction model described in the report. When running on 64 processors, the new MPI/OpenMP hybrid version has a wall clock speed up of 22 times compared to the current single processor version. When using 88 processors, a

  15. Image reconstruction of single photon emission computed tomography (SPECT) on a pebble bed reactor (PBR) using expectation maximization and exact inversion algorithms: Comparison study by means of numerical phantom

    Science.gov (United States)

    Razali, Azhani Mohd; Abdullah, Jaafar

    2015-04-01

    Single Photon Emission Computed Tomography (SPECT) is a well-known imaging technique used in medical application, and it is part of medical imaging modalities that made the diagnosis and treatment of disease possible. However, SPECT technique is not only limited to the medical sector. Many works are carried out to adapt the same concept by using high-energy photon emission to diagnose process malfunctions in critical industrial systems such as in chemical reaction engineering research laboratories, as well as in oil and gas, petrochemical and petrochemical refining industries. Motivated by vast applications of SPECT technique, this work attempts to study the application of SPECT on a Pebble Bed Reactor (PBR) using numerical phantom of pebbles inside the PBR core. From the cross-sectional images obtained from SPECT, the behavior of pebbles inside the core can be analyzed for further improvement of the PBR design. As the quality of the reconstructed image is largely dependent on the algorithm used, this work aims to compare two image reconstruction algorithms for SPECT, namely the Expectation Maximization Algorithm and the Exact Inversion Formula. The results obtained from the Exact Inversion Formula showed better image contrast and sharpness, and shorter computational time compared to the Expectation Maximization Algorithm.

  16. Image reconstruction of single photon emission computed tomography (SPECT) on a pebble bed reactor (PBR) using expectation maximization and exact inversion algorithms: Comparison study by means of numerical phantom

    Energy Technology Data Exchange (ETDEWEB)

    Razali, Azhani Mohd, E-mail: azhani@nuclearmalaysia.gov.my; Abdullah, Jaafar, E-mail: jaafar@nuclearmalaysia.gov.my [Plant Assessment Technology (PAT) Group, Industrial Technology Division, Malaysian Nuclear Agency, Bangi, 43000 Kajang (Malaysia)

    2015-04-29

    Single Photon Emission Computed Tomography (SPECT) is a well-known imaging technique used in medical application, and it is part of medical imaging modalities that made the diagnosis and treatment of disease possible. However, SPECT technique is not only limited to the medical sector. Many works are carried out to adapt the same concept by using high-energy photon emission to diagnose process malfunctions in critical industrial systems such as in chemical reaction engineering research laboratories, as well as in oil and gas, petrochemical and petrochemical refining industries. Motivated by vast applications of SPECT technique, this work attempts to study the application of SPECT on a Pebble Bed Reactor (PBR) using numerical phantom of pebbles inside the PBR core. From the cross-sectional images obtained from SPECT, the behavior of pebbles inside the core can be analyzed for further improvement of the PBR design. As the quality of the reconstructed image is largely dependent on the algorithm used, this work aims to compare two image reconstruction algorithms for SPECT, namely the Expectation Maximization Algorithm and the Exact Inversion Formula. The results obtained from the Exact Inversion Formula showed better image contrast and sharpness, and shorter computational time compared to the Expectation Maximization Algorithm.

  17. Image reconstruction of single photon emission computed tomography (SPECT) on a pebble bed reactor (PBR) using expectation maximization and exact inversion algorithms: Comparison study by means of numerical phantom

    International Nuclear Information System (INIS)

    Single Photon Emission Computed Tomography (SPECT) is a well-known imaging technique used in medical application, and it is part of medical imaging modalities that made the diagnosis and treatment of disease possible. However, SPECT technique is not only limited to the medical sector. Many works are carried out to adapt the same concept by using high-energy photon emission to diagnose process malfunctions in critical industrial systems such as in chemical reaction engineering research laboratories, as well as in oil and gas, petrochemical and petrochemical refining industries. Motivated by vast applications of SPECT technique, this work attempts to study the application of SPECT on a Pebble Bed Reactor (PBR) using numerical phantom of pebbles inside the PBR core. From the cross-sectional images obtained from SPECT, the behavior of pebbles inside the core can be analyzed for further improvement of the PBR design. As the quality of the reconstructed image is largely dependent on the algorithm used, this work aims to compare two image reconstruction algorithms for SPECT, namely the Expectation Maximization Algorithm and the Exact Inversion Formula. The results obtained from the Exact Inversion Formula showed better image contrast and sharpness, and shorter computational time compared to the Expectation Maximization Algorithm

  18. 球床高温气冷堆闭式循环特性%Characteristics of closed fuel cycles in the pebble bed high temperature gas cooled reactor

    Institute of Scientific and Technical Information of China (English)

    位金锋; 孙玉良; 李富

    2012-01-01

    The reuse of uranium and plutonium from high temperature gas-cooled reactor(HTGR) spent fuel will improve resource usage and minimize waste.The characteristics of different closed fuel cycles were studied here for uranium and plutonium recycled from 250 MWth high-temperature gas-cooled reactor pebble-bed-module(HTR-PM) spent fuel from a U-Pu fueled core.PuO2 and MOX fuel elements using recycled plutonium and uranium were then used in new PuO2 or MOX fueled cores with the same geometry as the original reactor.PuO2 from LWR spent fuel was also evaluated.The characteristics of the fuel utilization and transuranic incineration in these closed fuel cycles were studied with the VSOP program.The natural uranium utilization closed fuel for these closed fuel cycle is increased by 6%,8% and 20%,while the plutonium burn rates are 40%,41% and 63%,respectively.Thus,these HTGR closed fuel cycles can effectively burn plutonium isotopes and increase natural uranium utilization.%从提高天然铀利用率和改进废物管理方面考虑,研究球床高温气冷堆乏燃料中铀钚的再利用和不同闭式燃料循环的特性。在250MW热功率球床模块式高温气冷堆示范电站铀钚循环的乏燃料中提取铀和钚为核燃料,设计了PuO2和混合氧化物(MOX)燃料元件,将新设计的燃料元件重新装入与示范电站有同样结构和尺寸的堆芯,分别形成纯钚燃料循环和MOX燃料循环。还研究了基于轻水堆级钚的燃料循环。采用了高温气冷堆物理设计程序VSOP,研究了高温气冷堆不同闭式循环的燃料利用和超铀元素焚烧特性。不同闭式循环钚消耗率分别为50%、46%和71%,天然铀的电利用率分别提高了6%、8%和20%。结果表明:高温气冷堆闭式燃料循环能有效焚烧钚同位素,适度提高天然铀的利用率。

  19. 球床式高温气冷堆的余热不确定性分析%Uncertainty Analysis on Decay Heat of Pebble-bed High Temperature Gas-cooled Reactor

    Institute of Scientific and Technical Information of China (English)

    贠相羽; 郑艳华; 经荥清; 李富

    2013-01-01

    反应堆在停堆后相当长时间内仍具有较高的剩余发热是核电站的重要特性,也是核电站安全分析的关键.因此,对反应堆余热及其不确定性进行分析,对于合理设计余热排出系统、研究论证燃料元件在事故后的安全特性等均具有重要意义.本工作结合德国针对球床式高温气冷堆制定的余热计算标准,介绍了球床式高温气冷堆剩余发热及其不确定性的计算方法,并结合200 MWe球床模块式高温气冷堆示范工程(HTR-PM)的初步物理设计,对长期运行在满功率平衡堆芯状态下的反应堆停堆后的余热及其不确定性进行了计算分析,为进一步的事故分析提供依据.%The large amount of decay heat in a quite long time after reactor shutdown,which is an important characteristic of the nuclear power plants,should be considered seriously during the safety analysis.Therefore,the study on the decay heat and its uncertainty analysis play an important role in the design of decay heat removal system,as well as in the safety verification of the fuel element during the accident.In referenced to the standard of Germany entitled "Decay Heat Power in Nuclear Fuels of Hightemperature Reactors with Spherical Fuel Elements" especially for pebble-bed high temperature gas-cooled reactor (HTGR),the calculation method of decay heat and its uncertainty of pebble-bed HTGR were introduced.On the basis of the preliminary physical design of Chinese 200 MWe high temperature gas-cooled reactor pebble-bed module (HTR-PM),the decay heat and its uncertainty after reactor shutdown from long-term operation at rated power were analyzed,so as to provide a basis for further accident analysis.

  20. Neutronics R and D efforts in support of the European breeder blanket development programme

    International Nuclear Information System (INIS)

    The EU fusion technology programme considers two blanket development lines, the Helium-Cooled Pebble Bed (HCPB) blanket with Lithium ceramics pebbles as breeder material and beryllium pebbles as neutron multiplier, and the Helium-Cooled Lithium-Lead (HCLL) blanket with the Pb-Li eutectic alloy acting both as breeder and neutron multiplier. The long-term strategy aims at providing validated engineering designs of breeder blankets for a fusion power demonstration reactor (DEMO). As an important intermediate step, the breeder blankets need to be tested in a real fusion environment as provided by ITER. HCPB and HCLL Test Blanket Modules (TBM) have been accordingly designed for tests in dedicated ITER blanket ports. The nuclear design and performance of the breeder blanket modules rely on the results provided by neutronics design calculations. Validated computational tools and qualified nuclear data are required for high prediction accuracies including reliable uncertainty assessments. Complementary to the application of established standard tools and data for design analysis, a dedicated neutronics R and D effort is therefore conducted in the EU. This includes the development of dedicated computational tools, the generation of high quality nuclear data and their validation through integral experiments. The recent neutronic design efforts have been devoted to the European DEMO reactor study comprising (i) Monte Carlo based pre-analysis for the dimensioning of the shielding system, (ii) the generation of a generic CAD based Monte Carlo geometry model, and (iii) performance analysis for HCLL and HCPB based DEMO variants. The recent focus of the validation effort is on neutronics TBM mock-up experiments. The first experiment of this kind was performed on a TBM mock-up of the HCPB breeder blanket. The follow-up experiment on a neutronics HCLL TBM mock-up is currently under preparation. Computational pre-analysis were performed to optimise the design of the mock

  1. Analysis of the effect of random errors in burnup measurements in a pebble bed HTR%球床式高温气冷堆燃耗测量随机误差的影响分析

    Institute of Scientific and Technical Information of China (English)

    郝琛; 李富

    2013-01-01

    The fuel pebbles pass through the reactor core several times in a pebble bed high temperature gas cooled reactor (HTR).The burnup of the fuel balls is measured after they are discharged from the core bottom by a burnup measurement device.Those whose burnup does not reach the threshold are returned to the top of the core to pass through the core again,with the others transferred to the spent fuel storage tank.However,these random errors in the burnup measurement affect the fuel cycle.The MCPHS code,which is based on the Monte Carlo method,was used to analyze the characteristics of the pebble bed flow to simulate the random error in burnup measurement and its effect on the fuel cycle.The results show that the average discharge burnup and the distribution of the burnup in the core are not sensitive to the random error,while the distribution of the discharge burnup and the maximum and minimum discharge burnups are particularly sensitive to the random error in the burnup measurement.%球床式高温气冷堆燃料球多次通过堆芯,卸出堆芯的燃料球将由燃耗测量装置测量其燃耗,达到设定阈值的将按乏燃料处理,否则将返回堆芯继续裂变发热.而燃耗测量会具有随机误差,从而可能对燃料循环过程产生影响.该文改进了球床高温气冷堆燃料球运行历史的Monte Carlo模拟程序MCPHS,对燃耗测量的随机误差进行了模拟,对燃料循环过程的影响进行了分析.结果表明卸料燃耗均值、燃料球通过堆芯次数均值、堆芯燃耗分布对于燃耗测量误差并不敏感,而燃料球卸料燃耗分布、卸料燃耗最大值和最小值及燃料球通过堆芯最大值和最小值对于燃耗测量误差很敏感.

  2. Tritium recovery from helium purge stream of solid breeder blanket by cryogenic molecular sieve bed. 2. Regeneration operation of cryogenic molecular sieve bed

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Yoshinori; Enoeda, Mikio; Nishi, Masataka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Regeneration operation is a very important operation, because it is the most influential factor for deciding the net operation cycle time and the minimum dimension of Cryogenic Molecular Sieve Bed (CMSB). However, the experimental data of CMSB regeneration operation was not so sufficient that even the optimum regeneration procedure could not be decided yet. This work was focused on getting the primary information about various regeneration procedures. (author)

  3. Design and technology development of solid breeder blanket cooled by supercritical water in Japan

    International Nuclear Information System (INIS)

    successfully fabricated. It withstood the high heat flux test at 2.7 MW m-2. Also, a correlation parameter of the Li2TiO3 pebble bed made by the sol-gel method was verified by measurement of the thermal conductivity of the breeder pebble bed, which is one of the most important design data. (author)

  4. "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.

  5. Fusion breeder

    International Nuclear Information System (INIS)

    The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the US fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the US fusion program and the US nuclear energy program. The purpose of this paper is to suggest this policy change be made and tell why it should be made, and to outline specific research and development goals so that the fusion breeder will be developed in time to meet fissile fuel needs

  6. Studies on tritium breeding ratio for solid breeder blanket cooled by pressurized water through nuclear and thermal analyses

    International Nuclear Information System (INIS)

    Japan Atomic Energy Agency (JAEA) has been performing the research, development and design of blankets with water-cooled solid breeder for fusion power plant as a leading institute in Japan, according to the long-term R and D program established by the Fusion Council in 1999. For our design, pebbles of a ceramic tritium breeder (Li2TiO3) and a beryllium neutron multiplier (Be) are packed in the constitutive layer structures of a test blanket module (TBM) for ITER. These reports are results of one-dimensional nuclear and thermal analyses on the TBM emphasizing on optimized configuration of the breeder and multiplier layers. Taking into account increment of TBR, the radial widths of the breeder and multiplier layers are optimized. The main results of our study are as follows: (1) In multilayered structures of pebble beds, existence of the peak of the TBR was revealed within the range of the volume ratio R=V(Be)/V(Li2TiO3)=4-5. (2) In the case of optimized layer structure for the single packing, a layer of Be was set to be the two layers behind a layer of Li2TiO3. The R became available for staying in the range of R=4-5. Consequently, the TBR respectively increased by 2.0%, 3.2% and 4.0% with 7.5%(nature), 40% and 90% of enrichment of 6Li compared to TBR of TBM in which the layers of Be and Li2TiO3 were interlaminated. This database of TBR for optimized layer structure contributes to the estimation of TBR at the design stage of the TBM and demonstration blanket aimed to strengthen the commercial competitiveness and technical feasibility. (author)

  7. Set-up of a pre-test mock-up experiment in preparation for the HCPB Breeder Unit mock-up experimental campaign

    International Nuclear Information System (INIS)

    Highlights: ► As preparation for the HCPB-TBM Breeder Unit out-of-pile testing campaign, a pre-test experiment (PREMUX) has been prepared and described. ► A new heater system based on a wire heater matrix has been developed for imitating the neutronic volumetric heating and it is compared with the conventional plate heaters. ► The test section is described and preliminary thermal results with the available models are presented and are to be benchmarked with PREMUX. ► The PREMUX integration in the air cooling loop L-STAR/LL in the Karlsruhe Institute for Technology is shown and future steps are discussed. -- Abstract: The complexity of the experimental set-up for testing a full-scaled Breeder Unit (BU) mock-up for the European Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) has motivated to build a pre-test mock-up experiment (PREMUX) consisting of a slice of the BU in the Li4SiO4 region. This pre-test aims at verifying the feasibility of the methods to be used for the subsequent testing of the full-scaled BU mock-up. Key parameters needed for the modeling of the breeder material is also to be determined by the Hot Wire Method (HWM). The modeling tools for the thermo-mechanics of the pebble beds and for the mock-up structure are to be calibrated and validated as well. This paper presents the setting-up of PREMUX in the L-STAR/LL facility at the Karlsruhe Institute of Technology. A key requirement of the experiments is to mimic the neutronic volumetric heating. A new heater concept is discussed and compared to several conventional heater configurations with respect to the estimated temperature distribution in the pebble beds. The design and integration of the thermocouple system in the heater matrix and pebble beds is also described, as well as other key aspects of the mock-up (dimensions, layout, cooling system, purge gas line, boundary conditions and integration in the test facility). The adequacy of these methods for the full-scaled BU mock-up is

  8. R and D activities on helium cooled solid breeder TBM in Korea

    International Nuclear Information System (INIS)

    R and D activities currently being undertaken for HCSB TBM include joining technologies of structural material, breeder and reflector pebble material development, the effect of TBM ferritic-martensitic steel on the ripple of toroidal magnetic field, and ceramic coating on graphite pebble. The HIP joining performance of FM steel is evaluated. Lithium ceramic breeder and graphite reflector pebble fabrication methods are under development using special fabrication process, and the initial characteristics of the pebbles are assessed. Silicon carbide coating on graphite pebble is also investigated and its preliminary results are mentioned. Finally, an accurate evaluation of the effect of TBM and ferromagnetic inserts on magnetic field are implemented. The current results of these R and D issues are addressed in this paper.

  9. 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)

  10. Exploring new coolants for nuclear breeder reactors

    International Nuclear Information System (INIS)

    Breeder reactors are considered a unique tool for fully exploiting natural nuclear resources. In current Light Water Reactors (LWR), only 0.5% of the primary energy contained in the nuclei removed from a mine is converted into useful heat. The rest remains in the depleted uranium or spent fuel. The need to improve resource-efficiency has stimulated interest in Fast-Reactor-based fuel cycles, which can exploit a much higher fraction of the energy content of mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles also offer several potential advantages over a uranium fuel cycle. The coolant initially selected for most of the FBR programs launched in the 1960s was sodium, which is still considered the best candidate for these reactors. However, Na-cooled FBRs have a positive void reactivity coefficient. Among other factors, this fundamental drawback has resulted in the canceled deployment of these reactors. Therefore, it seems reasonable to explore new options for breeder coolants. In this paper, a proposal is presented for a new molten salt (F2Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would require an extensive R and D program, this paper presents the very appealing properties of this salt when using a specific type of fuel that is similar to that of pebble bed reactors. The F2Be concept was studied over a typical MOX composition and extended to a thorium-based cycle. The general analysis took into account the requirements for criticality (opening the option of hybrid subcritical systems); the requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window was found in the definition of a F2Be cooled reactor where the safety requirement was met, unlike for molten metal-cooled reactors, which always have positive void reactivity coefficients

  11. 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.

  12. Fabrication tests of Li2TiO3 pebbles by direct wet process

    International Nuclear Information System (INIS)

    Lithium titanate (Li2TiO3) pebbles are considered to be the candidate material of the tritium breeders for fusion reactor from a point of good tritium recovery, chemical stability, etc. The direct wet process that Li2TiO3 pebbles were fabricated from the Li2TiO3 solution directly was proposed. In this study, pebble fabrication tests by the direct wet process were performed. The results from the preliminary test were as follows: 1) 100% Li2TiO3 powder could be dissolved when the holding time at more than 60 C was longer. 2) Good gel shape was maintained by dropping the Li2TiO3 condensed solution liquid in acetone. 3) Adjustment of a solution influenced the cracking rate of the Li2TiO3 pebble surface. Additionally, the solvent exchange was effective to decrease the crack of Li2TiO3 pebble surface and to improve the density of Li2TiO3 pebbles. It was clear that Li2TiO3 pebbles could be fabricated by the direct wet process and the pebbles with 5 μm grain and uniform structure were obtained. (orig.)

  13. Status of advanced tritium breeder development for DEMO in the broader approach activities in Japan

    International Nuclear Information System (INIS)

    DEMO reactors require '6Li-enriched ceramic tritium breeders' which have high tritium breeding ratios (TBRs) in the blanket designs of both EU and JA. Both parties have been promoting the development of fabrication technologies of Li2TiO3 pebbles and of Li4SiO4 pebbles including the reprocessing. However, the fabrication techniques of tritium breeders pebbles have not been established for large quantities. Therefore, these parties launch a collaborative project on scaleable and reliable production routes of advanced tritium breeders. In addition, this project aims to develop fabrication techniques allowing effective reprocessing of 6Li. The development of the production and 6Li reprocessing techniques includes preliminary fabrication tests of breeder pebbles, reprocessing of lithium, and suitable out-of-pile characterizations. The R and D on the fabrication technologies of the advanced tritium breeders and the characterization of developed materials has been started between the EU and Japan in the DEMO R and D of the International Fusion Energy Research Centre (IFERC) project as a part of the Broader Approach activities from 2007 to 2016. The equipment for production of advanced breeder pebbles is planned will be installed in the DEMO R and D building at Rokkasho, Japan. The design work in this facility was carried out. The specifications of the pebble production apparatuses and related equipment in this facility were fixed, and the basic data of these apparatuses was obtained. In this design work, the preliminary investigations of the dissolution and purification process of tritium breeders were carried out. From the results of the preliminary investigations, lithium resources of 90% above were recovered by the aqueous dissolving methods using HNO3 and H2O2. The removal efficiency of 60Co by the addition in the dissolved solutions of lithium ceramics were 97-99.9% above using activated carbon impregnated with 8-hydroxyquinolinol. In this report, preparation status

  14. Experimental and numerical investigations of heat transfer in the first wall of Helium-Cooled-Pebble-Bed Test Blanket Module – Part 1: Presentation of test section and 3D CFD model

    International Nuclear Information System (INIS)

    Highlights: • Design of the test section for investigation of heat transfer in the first wall is presented. • Manufacturing details and providing of operational ready mock-up are given. • Corresponding 3D CFD model of the test section is described. - Abstract: This paper deals with cooling of the first wall of Helium-Cooled-Pebble-Bed Test Blanket Module (HCPB TBM) for ITER. The first wall cooling is an important investigation issue due to an extreme asymmetry of heat loads: heat flux on the plasma facing side is several times stronger than the one on the side which faces breeding units. Our preliminary 3D CFD analysis revealed that under such conditions the heat transfer coefficient is significantly lower than predicted by common heat transfer correlations (see Ilić et al., 2006). For an experimental validation of these results HETRA (HEat TRAnsfer) test section has been designed and built at the Institute for Neutron Physics and Reactor Technology in Karlsruhe Institute of Technology. The HETRA test section involves in full scale one U-pass of the cooling channel in the first wall of HCPB TBM Version 1.1 (see Meyder et al., 2005). The HCPB TBM relevant experimental conditions have been provided: test channel made of Eurofer steel, helium coolant at pressure of 8 MPa and inlet temperature of 300 °C and heat flux of 270 kW/m2 at the channel surface representing plasma facing side of the first wall. Test channels with hydraulically smooth and with hydraulically rough walls have been built. At each test channel the temperature of Eurofer walls has been measured at ∼60 positions. For numerical investigations the 3D CFD modelling with the code STAR CD has been applied. This paper is the first report on this study and presents the development of the test section and of the 3D CFD model. The analyses of the obtained experimental and computational results are presented in the second report (see Ilić et al., 2014)

  15. Monte Carlo criticality calculation for Pebble-type HTR-PROTEUS core

    International Nuclear Information System (INIS)

    These days, pebble-bed and other High-Temperature Gas-cooled Reactor (HTGR) designs are once again in vogue in connection with hydrogen production. In this study, as a part of establishing Monte Carlo computation system for HTGR core analysis, some criticality calculations for pebble-type HTGR were carried out using MCNP code. Firstly, the pebble-bed cores of HTR-PROTEUS critical facility in Swiss were selected for the benchmark model, and, after the detailed MCNP modeling of the whole facility, criticality calculations were performed. It was also investigated the homogenization effect of TRISO fuel on criticality

  16. Evaluation of tritium release properties of advanced tritium breeders

    International Nuclear Information System (INIS)

    Demonstration power plant (DEMO) fusion reactors require advanced tritium breeders with high thermal stability. Lithium titanate (Li2TiO3) advanced tritium breeders with excess Li (Li2+xTiO3+y) are stable in a reducing atmosphere at high temperatures. Although the tritium release properties of tritium breeders are documented in databases for DEMO blanket design, no in situ examination under fusion neutron (DT neutron) irradiation has been performed. In this study, a preliminary examination of the tritium release properties of advanced tritium breeders was performed, and DT neutron irradiation experiments were performed at the fusion neutronics source (FNS) facility in JAEA. Considering the tritium release characteristics, the optimum grain size after sintering is <5 μm. From the results of the optimization of granulation conditions, prototype Li2+xTiO3+y pebbles with optimum grain size (<5 μm) were successfully fabricated. The Li2+xTiO3+y pebbles exhibited good tritium release properties similar to the Li2TiO3 pebbles. In particular, the released amount of HT gas for easier tritium handling was higher than that of HTO water. (authors)

  17. Inspection sensor for pebble bed reactors

    International Nuclear Information System (INIS)

    In order to inspect the inside surface of the reflector a lance-shaped sensor is used, in the head of which the objective lens of a periscope is situated. So that unevenness of the reflector surface can be compensated for, the head of the sensor is made movable via a rotating joint and by spring force relative to the axial direction of the shaft of the lance. (DG)

  18. Tritium release from highly neutron irradiated constrained and unconstrained beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Chakin, V., E-mail: vladimir.chakin@kit.edu; Rolli, R.; Vladimirov, P.; Moeslang, A.

    2015-06-15

    Highlights: • For the irradiated constrained beryllium pebbles, the tritium release occurs easier than for the unconstrained ones. • Tritium retention in the irradiated constrained and unconstrained beryllium pebbles decreases with increasing irradiation temperature. • Formation of sub-grains in the constrained beryllium pebbles facilitate the open porosity network formation. - Abstract: Beryllium is the reference neutron multiplier material in the Helium Cooled Pebble Bed (HCPB) breeding blanket of fusion power plants. Significant tritium inventory accumulated in beryllium as a result of neutron-induced transmutations could become a safety issue for the operation of such blankets as well as for the nuclear waste utilization. To provide a related materials database, a neutron irradiation campaign of beryllium pebbles with diameters of 0.5 and 1 mm at 686–1006 K, the HIDOBE-01 experiment, has been performed in the HFR in Petten, the Netherlands, producing up to 3020 appm helium and 298 appm tritium. Thermal desorption tests of irradiated unconstrained and constrained beryllium pebbles were performed in a purge gas flow using a quadrupole mass-spectrometer (QMS) and an ionization chamber. Compared to unconstrained pebbles, constrained beryllium pebbles have an enhanced tritium release at all temperatures investigated. Small elongated sub-grains formed under irradiation in the constrained pebbles promote formation of numerous channels for facilitated tritium release.

  19. Tritium release from highly neutron irradiated constrained and unconstrained beryllium pebbles

    International Nuclear Information System (INIS)

    Highlights: • For the irradiated constrained beryllium pebbles, the tritium release occurs easier than for the unconstrained ones. • Tritium retention in the irradiated constrained and unconstrained beryllium pebbles decreases with increasing irradiation temperature. • Formation of sub-grains in the constrained beryllium pebbles facilitate the open porosity network formation. - Abstract: Beryllium is the reference neutron multiplier material in the Helium Cooled Pebble Bed (HCPB) breeding blanket of fusion power plants. Significant tritium inventory accumulated in beryllium as a result of neutron-induced transmutations could become a safety issue for the operation of such blankets as well as for the nuclear waste utilization. To provide a related materials database, a neutron irradiation campaign of beryllium pebbles with diameters of 0.5 and 1 mm at 686–1006 K, the HIDOBE-01 experiment, has been performed in the HFR in Petten, the Netherlands, producing up to 3020 appm helium and 298 appm tritium. Thermal desorption tests of irradiated unconstrained and constrained beryllium pebbles were performed in a purge gas flow using a quadrupole mass-spectrometer (QMS) and an ionization chamber. Compared to unconstrained pebbles, constrained beryllium pebbles have an enhanced tritium release at all temperatures investigated. Small elongated sub-grains formed under irradiation in the constrained pebbles promote formation of numerous channels for facilitated tritium release

  20. Stepped-anneal helium release in 1-mm beryllium pebbles from COBRA-1A2

    Energy Technology Data Exchange (ETDEWEB)

    Oliver, B.M. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    Stepped-anneal helium release measurements on two sets of fifteen beryllium pebbles irradiated in the Experimental Breeder Reactor-II (EBR-II) at Argonne National Laboratory-West (ANL-w), are reported. The purpose of the measurements was to determine the helium release characteristics of the beryllium using larger sample sizes and longer anneal times relative to earlier measurements. Sequential helium analyses were conducted over a narrower temperature range from approximately 800 C to 1100 C in 100 C increments, but with longer anneal time periods. To allow for overnight and unattended operation, a temperature controller and associated circuitry were added to the experimental setup. Observed helium release was nonlinear with time at each temperature interval, with each step being generally characterized by an initial release rate followed by a slowing of the rate over time. Sample Be-C03 showed a leveling off in the helium release after approximately 3 hours at a temperature of 890 C. Sample Be-D03, on the other hand, showed a leveling off only after {approximately}12 to 24 hours at a temperature of 1100 C. This trend is consistent with that observed in earlier measurements on single microspheres from the same two beryllium lots. None of the lower temperature steps showed any leveling off of the helium release. Relative to the total helium concentrations measured earlier, the total helium releases observed here represent approximately 80% and 92% of the estimated total helium in the C03 and D03 samples, respectively.

  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. Solid breeder blanket concepts

    International Nuclear Information System (INIS)

    An investigation is made of a mechanical concept for the blanket with solid breeders in view of the possible adaptation to power reactor. A special arrangement of the multiplier and breeder materials is developed to permit a further neutronic optimisation

  3. 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.

  4. TEM study of impurity segregations in beryllium pebbles

    Science.gov (United States)

    Klimenkov, M.; Chakin, V.; Moeslang, A.; Rolli, R.

    2014-12-01

    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.

  5. 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

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

    International Nuclear Information System (INIS)

    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-PREX facility uses novel digital

  7. Thermohydraulics design and thermomechanics analysis of two European breeder blanket concepts for DEMO. Pt. 1 and Pt. 2. Pt. 1: BOT helium cooled solid breeding blanket. Pt. 2: Dual coolant self-cooled liquid metal blanket

    International Nuclear Information System (INIS)

    Two different breeding blanket concepts are being elaborated at Forschungszentrum Karlsruhe within the framework of the DEMO breeding blanket development, the concept of a helium cooled solid breeding blanket and the concept of a self-cooled liquid metal blanket. The breeder material used in the first concept is Li4SiO4 as a pebble bed arranged separate from the beryllium pebble bed, which serves as multiplier. The breeder material zone is cooled by several toroidally-radially configurated helium cooling plates which, at the same time, act as reinforcements of the blanket structures. In the liquid metal blanket concept lead-lithium is used both as the breeder material and the coolant. It flows at low velocity in poloidal direction downwards and back in the blanket front zone. In both concepts the First Wall is cooled by helium gas. This report deals with the thermohydraulics design and thermomechanics analysis of the two blanket concepts. The performance data derived from the Monte-Carlo computations serve as a basis for the design calculations. The coolant inlet and outlet temperatures are chosen with the design criteria and the economics aspects taken into account. Uniform temperature distribution in the blanket structures can be achieved by suitable branching and routing of the coolant flows which contributes to reducing decisively the thermal stress. The computations were made using the ABAQUS computer code. The results obtained of the stresses have been evaluated using the ASME code. It can be demonstrated that all maximum values of temperature and stress are below the admissible limit. (orig.)

  8. Breeder Reprocessing Engineering Test

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, C.A.; Meacham, S.A.

    1984-01-01

    The Breeder Reprocessing Engineering Test (BRET) is a developmental activity of the US Department of Energy to demonstrate breeder fuel reprocessing technology while closing the fuel cycle for the Fast Flux Test Facility (FFTF). It will be installed in the existing Fuels and Materials Examination Facility (FMEF) at the Hanford Site near Richland, Washington, The major objectives of BRET are: (1) close the US breeder fuel cycle; (2) develop and demonstrate reprocessing technology and systems for breeder fuel; (3) provide an integrated test of breeder reactor fuel cycle technology - rprocessing, safeguards, and waste management. BRET is a joint effort between the Westinghouse Hanford Company and Oak Ridge National Laboratory. 3 references, 2 figures.

  9. Analysis of the HCPB breeder blanket bock-up experiment for ITER using SUSD3D code

    International Nuclear Information System (INIS)

    In order to validate new nuclear cross-section evaluations, method development and design of the helium-cooled pebble bed (HCPB) test blanket module of ITER a benchmark experiment was performed this year at the Frascati Neutron Generator (FNG) in the scope of the EFF (European Fusion File) project in Europe. The objective of this experiment is to study the tritium breeding ratio and other nuclear quantities in a breeder blanket in order to establish and improve the quality of related JEFF nuclear data. The experiment consists of a metallic beryllium set-up with two double layers of breeder material (Li2CO3 powder). The reaction rate measurements include the Li2CO3 pellets (tritium breeding ratio), activation foils, and neutron and gamma spectrometers inserted at several axial and lateral locations in the block. Our task is to perform the deterministic transport, and cross section sensitivity and uncertainty analysis. The role of the cross-section sensitivity and uncertainty analysis is to optimise the design of the benchmark, and to assist in the interpretation of the measurement results. The paper presents the pre- and post- analysis of the benchmark experiment. (author)

  10. Neutronics Experiment on A HCPB Breeder Blanket Mock-Up

    International Nuclear Information System (INIS)

    A neutronics experiment has been performed in the frame of European Fusion Technology Program on a mock-up of the EU Test Blanket Module (TBM), Helium Cooled Pebble Bed (HCPB) concept, with the objective to validate the capability of nuclear data to predict nuclear responses, such as the tritium production rate (TPR), with qualified uncertainties. The experiment has been carried out at the FNG 14-MeV neutron source in collaboration between ENEA, Technische Universitaet Dresden, Forschungszentrum Karlsruhe, J. Stefan Institute Ljubljana and with the participation of JAEA. The mock-up, designed in such a way to replicate all relevant nuclear features of the TBM-HCPB, consisted of a steel box containing beryllium block and two intermediate steel cassettes, filled with of Li2CO3 powder, replicating the breeder insert main characteristics: radial thickness, distance between ceramic layers, thickness of ceramic layers and of steel walls. In the experiment, the TPR has been measured using Li2CO3 pellets at various depths at two symmetrical positions at each depth, one in the upper and one in the lower cassette. Twelve pellets were used at each position to determine the TPR profile through the cassette. Three independent measurements were performed by ENEA, TUD/VKTA and JAEA. The neutron flux in the beryllium layer was measured as well using activation foils. The measured tritium production in the TBM (E) was compared with the same quantity (C) calculated by the MCNP.4c using a very detailed model of the experimental set up, and using neutron cross sections from the European Fusion File (EFF ver.3.1) and from the Fusion Evaluated Nuclear Data Library (FENDL ver. 2.1, ITER reference neutron library). C/E ratios were obtained with a total uncertainty on the C/E comparison less than 9% (2 s). A sensitivity and uncertainty analysis has also been performed to evaluate the calculation uncertainty due to the uncertainty on neutron cross sections. The results of such analysis

  11. RAZVOJ APLIKACIJ ZA PAMETNO URO PEBBLE

    OpenAIRE

    Kranjc, Denis

    2015-01-01

    V diplomski nalogi predstavljamo razvoj aplikacij za pametno uro Pebble in razvoj aplikacij za pametni telefon Android, ki komunicira z uro preko Bluetooth povezave. Pri razvoju smo uporabili razvojno okolje CloudPebble, programski razvojni paket Pebble SDK, razvojno okolje Android Studio in javansko knjižnico PebbleKit. Aplikacije za pametno uro smo razvijali v programskem jeziku C, aplikacije Android pa v programskem jeziku Java. Rezultat diplomskega dela je osem razvitih različnih aplikaci...

  12. TEM study of beryllium pebbles after neutron irradiation up to 3000 appm helium production

    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)

    2013-11-15

    Beryllium is planned to be used as a neutron multiplier in the Helium Cooled Pebble Bed (HCPB) European concept of a breeding blanket of DEMO. In order to evaluate the irradiation performance, individual pebbles and constrained pebble beds were neutron irradiated at temperatures typical for fusion blanket. Beryllium pebbles with a diameter of 1 mm produced by the Rotating Electrode Method were subjected to a TEM study after irradiation at the HFR, Petten, at temperatures of 686, 753, 861, and 968 K. The helium production in the pebbles was calculated in the range from 2090 to 3090 appm. Gas bubbles as disks of hexagonal shape were observed for all four irradiation temperatures. The disks were oriented in the (0 0 0 1) basal plane with a height directed along the [0 0 0 1] “c” axis. The average diameters of the bubbles increase from 7.5 to 80 nm with increasing irradiation temperature, the bulk densities accordingly decrease from 4.4 × 10{sup 22} to 3.8 × 10{sup 20} m{sup −3}. With increasing irradiation temperature, the swelling of the pebbles increases from 0.6% at 686 K up to 6.5% at 968 K.

  13. Evidence for Pebbles in Comets

    CERN Document Server

    Kretke, K A

    2015-01-01

    When the EPOXI spacecraft flew by Comet 103P/Hartley 2, it observed large particles floating around the comet nucleus. These particles are likely low-density, centimeter- to decimeter-sized clumps of ice and dust. While the origin of these objects remains somewhat mysterious, it is possible that they are giving us important information about the earliest stages of our Solar System's formation. Recent advancements in planet formation theory suggest that planetesimals (or cometestimals) may grow directly from the gravitational collapse of aerodynamically concentrated small particles, often referred to as "pebbles." Here we show that the particles observed in the coma of 103P are consistent with the sizes of pebbles expected to efficiently form planetesimals in the region that this comet likely formed, while smaller pebbles are may be expected in the majority of comets, whose chemistry is often indicative of formation in the colder, outer regions of the protoplanetary disk.

  14. Fast breeder reactor

    International Nuclear Information System (INIS)

    The fluid-cooled fast breeder reactor described includes an outer cylindrical boundary wall, a plurality of canless fuel elements and breeder material elements received within the boundary wall and being in an array therein forming a fissionable fuel zone and a breeder material zone coaxially surrounding the fissionable fuel zone, a coolant supply system for applying fluid coolant at uniform pressure to the entire cross section within the cylindrical boundary wall, and flow guide devices extending substantially horizontally and disposed at different levels one above the other within the breeder material zone which coaxially surrounds the fissionable fuel zone, means for elastically securing the flow guide devices at alternate levels within the breeder material to the boundary wall, the flow guide devices at the levels intermediate the alternate levels being spaced by an annular gap from the boundary wall. 7 claims, 7 drawing figures

  15. Embattled breeder reactor

    International Nuclear Information System (INIS)

    A commercial fuel-cloning machine, a nuclear breeder reactor, is yet to produce electricity in the United States. It is expensive in capital and fuel costs, its fuel that must be reprocessed can become a link to nuclear weapons manufacture, and its safety is no greater than conventional nuclear reactors. The breeder has had on-again/off-again administrative support from Washington. Opponents worry about escalating costs and failure to develop alternatives like solar energy. Proponents say fossil-fuel depletion will eventually force long-term renewable resources such as the breeder anyway. Some who share parts of both views oppose present policy regarding the Clinch River Breeder demonstration plant specifically. The correct choices on breeder concept development and commercialization will be known in 2050. 3 figures

  16. A Feasible DEMO Blanket Concept Based on Water Cooled Solid Breeder

    International Nuclear Information System (INIS)

    Full text: JAEA has conducted the conceptual design study of blanket for a fusion DEMO reactor SlimCS. Considering DEMO specific requirements, we place emphasis on a blanket concept with durability to severe irradiation, ease of fabrication for mass production, operation temperature of blanket materials, and maintainability using remote handling equipment. This paper present a promising concept satisfying these requirements, which is characterized by minimized welding lines near the front, a simplified blanket interior consisting of cooling tubes and a mixed pebble bed of breeder and neutron multiplier, and approximately the same outlet temperature for all blanket modules. Neutronics calculation indicated that the blanket satisfies a self-sufficient production of tritium. An important finding is that little decrease is seen in tritium breeding ratio even when the gap between neighboring blanket modules is as wide as 0.03 m. This means that blanket modules can be arranged with such a significant clearance gap without sacrifice of tritium production, which will facilitate the access of remote handling equipment for replacement of the blanket modules and improve the access of diagnostics. (author)

  17. Cold trapping of traces of tritiated water from the helium loops of a fusion breeder blanket

    International Nuclear Information System (INIS)

    The ITER Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) will comprise three helium loops designed for: tritium extraction from the breeder zone, heat removal, and purification of the coolant. The process step envisaged for tritium extraction as well as for coolant purification includes a cryogenic cold trap as main component for the removal of tritiated water vapour (mainly HTO, H2O). The concentrations of water in the gas streams are expected to be extremely small, i.e. of the order of 10 ppm by volume. In this paper, we describe first runs with a cold trap using helium as the carrier gas at flow rates of 0.1 and 1.0 m3/h. The range of water vapour concentration in the helium carrier gas was 0.5 to >200 ppmv. The experiments have demonstrated the ability of the cold trap to remove water vapour efficiently from the He stream down to concentrations of less than 0.02 ppmv when the inlet water concentration is in the range of 300-650 ppmv or higher

  18. Proceedings of the sixth international workshop on ceramic breeder blanket interactions

    International Nuclear Information System (INIS)

    This report is the Proceedings of ''the Sixth International Workshop on Ceramic Breeder Blanket Interactions'' which was held as a workshop on ceramic breeders under Annex II of IEA Implementing Agreement on a Programme of Research and Development on Fusion Materials, and Japan-US Workshop 97FT4-01. This workshop was held in Mito city, Japan on October 22-24, 1997. About forty experts from EU, Japan, USA, and Chile attended the workshop. The scope of the workshop included the following: 1) fabrication and characterization of ceramic breeders, 2) properties data for ceramic breeders, 3) tritium release characteristics, 4) modeling of tritium behavior, 5) irradiation effects on performance behavior, 6) blanket design and R and D requirements, 7) hydrogen behavior in materials, and 8) blanket system technology and structural materials. In the workshop, information exchange was performed for fabrication technology of ceramic breeder pebbles in EU and Japan, data of various properties of Li2TiO3, tritium release behavior of Li2TiO3 and Li2ZrO3 including tritium diffusion, modeling of tritium release from Li2ZrO3 in ITER condition, helium release behavior from Li2O, results of tritium release irradiation tests of Li4SiO4 pebbles in EXOTIC-7, R and D issues for ceramic breeders for ITER and DEMO blankets, etc. The 23 of the papers are indexed individually. (J.P.N.)

  19. Achievements of the water cooled solid breeder test blanket module of Japan to the milestones for installation in ITER

    International Nuclear Information System (INIS)

    As the primary candidate of ITER Test Blanket Module (TBM) to be tested under the leadership of Japan, Water Cooled Solid Breeder (WCSB) TBM is being developed. Six TBMs will be tested in ITER simultaneously, under the leadership of different countries. To ensure the installation of reliable TBMs, it is necessary to show feasibility on the TBM milestones for installation in ITER. This paper shows the recent achievements toward the milestones of ITER TBMs prior to the installation, that consist of design integration in ITER, module qualification and safety assessment. With respect to the design integration, it is necessary to show the consistency with ITER design on time with ITER design progress, targeting the detailed design final report in 2012. Structure design of the interfacing components between the WCSB TBM structure and the interfacing components (Common Frame and Backside Shielding) that are placed in a test port of ITER has been developed. The design work also consists of procedures of fabrication and replacement of TBM, the consistency with ITER port structure and TBM interface structure, and the layouts of the auxiliary systems of TBMs including the tritium extraction system and water cooling system. As for the module qualification, it is necessary to show fabrication capability and the integrity of prototypical size mockup in corresponding operation condition before the delivery of the TBM to ITER. A real scale first wall mock-up was successfully fabricated by using Hot Isostatic Pressing (HIP) method by structural material of reduced activation martensitic ferritic steel, F82H. High heat flux test with real cooling water condition is planned using this mock-up. Other essential R and Ds for the WCSB TBM also showed steady progress on investigation of mechanical behavior of breeder pebble beds, development of advanced breeder/multiplier pebble, neutron measurement technology for TBM and purge gas tritium recovery technology. As for safety milestones

  20. Conceptual design of solid breeder blanket system cooled by supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio; Akiba, Masato [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment] [and others

    2001-12-01

    This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The discussion of the Fusion Council in 1999 updated the assessment of the mission of DEMO blanket. Updated mission of the DEMO blanket is to be the prototype of the commercially competitive power plant. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. From such standing point, the conceptual design study was performed to determine the updated strategy and goal of the R and D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology. The DEMO blanket applies the solid breeder materials and supercritical water cooling. The product tritium is purged out by helium gas stream in the breeder region. In the breeder region, the pebble bed concept was applied to withstand instable cracking of the breeder and multiplier materials in high neutron irradiation and high temperature operation. Inlet temperature of the coolant is planned to be 280degC and final outlet temperature is 510degC to obtain high energy conversion efficiency up to 43%. Reduced activation ferritic steel, F82H and ODS ferritic steel were selected as the structural material. Lithium ceramics, Li{sub 2}TiO{sub 3} or Li{sub 2}O were selected as the breeder materials. Beryllium or its inter-metallic compound Be12Ti was selected as the neutron multiplier materials. Basic module structure was selected as the box type structure which enables the remote handling replacement of the module from in-vessel access. Dimension of the box is limited to 2 m x 2 m, or smaller, due to the dimension of the replacement port. In the supercritical water cooling, the high coolant temperature is the merit for

  1. Conceptual design of solid breeder blanket system cooled by supercritical water

    International Nuclear Information System (INIS)

    This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The discussion of the Fusion Council in 1999 updated the assessment of the mission of DEMO blanket. Updated mission of the DEMO blanket is to be the prototype of the commercially competitive power plant. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. From such standing point, the conceptual design study was performed to determine the updated strategy and goal of the R and D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology. The DEMO blanket applies the solid breeder materials and supercritical water cooling. The product tritium is purged out by helium gas stream in the breeder region. In the breeder region, the pebble bed concept was applied to withstand instable cracking of the breeder and multiplier materials in high neutron irradiation and high temperature operation. Inlet temperature of the coolant is planned to be 280degC and final outlet temperature is 510degC to obtain high energy conversion efficiency up to 43%. Reduced activation ferritic steel, F82H and ODS ferritic steel were selected as the structural material. Lithium ceramics, Li2TiO3 or Li2O were selected as the breeder materials. Beryllium or its inter-metallic compound Be12Ti was selected as the neutron multiplier materials. Basic module structure was selected as the box type structure which enables the remote handling replacement of the module from in-vessel access. Dimension of the box is limited to 2 m x 2 m, or smaller, due to the dimension of the replacement port. In the supercritical water cooling, the high coolant temperature is the merit for the energy

  2. Fast Breeder Reactor studies

    International Nuclear Information System (INIS)

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts

  3. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  4. Kinetics of tritium release from irradiated Li2TiO3 pebbles in out-of-pile TPD tests

    International Nuclear Information System (INIS)

    The rate of tritium release from Li2TiO3 pebbles was examined by post irradiation thermal desorption spectroscopy (the Temperature Programmed Desorption (TPD) method). Pre-treatments before and even after irradiation were found useful to gain insight on the behavior of these pebbles at different temperatures, as good spectrum de-convolution is achieved and kinetic parameters for the rate determining pseudo-first-order steps can be estimated. We show the results concerning Li2TiO3 pebbles bed specimens developed in the frame of the European fusion technology program

  5. A smoother pebble mathematical explorations

    CERN Document Server

    Benson, Donald C

    2003-01-01

    Introduction. I. BRIDGING THE GAP. 1. The Ancient Fractions. 2. Greek Gifts. 3. The Music of the Ratios. II. THE SHAPE OF THINGS. 4. Tubeland. 5. The Calculating Eye. III. THE GREAT ART. 6. Algebra Rules. 7. The Root of the Problem. 8. Symmetry Without Fear. 9. The Magic Mirror. IV. A SMOOTHER PEBBLE. 10. On the Shoulders of Giants. 11. Six-Minute Calculus. 12. Roller-Coaster Science. Glossary. References. Index

  6. EXOTIC-7: irradiation of ceramic breeder materials to high lithium burnup

    International Nuclear Information System (INIS)

    The EXOTIC-7 irradiation experiment in the high flux reactor (HFR) has been completed. Its aim has been to investigate the effects of high lithium-burnup on the mechanical stability and tritium release characteristics of candidate ceramic breeder materials, originating from the fusion programmes of CEA, FZK, ENEA, AECL and ECN. The tested ceramic breeder materials were pellets of Li2ZrO3, LiAlO2 and Li8ZrO6 and pebbles of Li4SiO4 and Li2ZrO3, with a variety of characteristics, like grain size and porosity. The test matrix provided the simultaneous irradiation of eight independent capsules with on-line tritium monitoring. Two capsules contained a mixture of Li4SiO4 and beryllium pebbles. The experimental design, sample loading and main irradiation parameters are described. Some PIE results and analysis of in-situ tritium release behaviour are presented. (orig.)

  7. Pebble red modular reactor - South Africa

    International Nuclear Information System (INIS)

    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)

  8. The fusion breeder

    International Nuclear Information System (INIS)

    The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the U.S. fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the U.S. fusion program and the U.S. nuclear energy program. There is wide agreement that many approaches will work and will produce fuel for five equal-sized LWRs, and some approach as many as 20 LWRs at electricity costs within 20% of those at today's price of uranium ($30/lb of U3O8). The blankets designed to suppress fissioning, called symbiotes, fusion fuel factories, or just fusion breeders, will have safety characteristics more like pure fusion reactors and will support as many as 15 equal power LWRs. The blankets designed to maximize fast fission of fertile material will have safety characteristics more like fission reactors and will support 5 LWRs. This author strongly recommends development of the fission suppressed blanket type, a point of view not agreed upon by everyone. There is, however, wide agreement that, to meet the market price for uranium which would result in LWR electricity within 20% of today's cost with either blanket type, fusion components can cost severalfold more than would be allowed for pure fusion to meet the goal of making electricity alone at 20% over today's fission costs. Also widely agreed is that the critical-pathitem for the fusion breeder is fusion development itself; however, development of fusion breeder specific items (blankets, fuel cycle) should be started now in order to have the fusion breeder by the time the rise in uranium prices forces other more costly choices

  9. Clinch River Breeder Reactor

    International Nuclear Information System (INIS)

    Mr. Baron says the administration's effort to terminate the Clinch River Breeder Reactor (CRBR) project is symptomatic; they have also placed restrictions on fusion, coal, solar, and other areas of energy development in which technological advances are held back in order to force conservation. Because the breeder reactor, unlike solar and fusion energy, is both economically and technically feasible, a demonstration plant is needed. The contentions that the CRBR design is obsolete, that its proposed size is inappropriate, or that plutonium can be diverted for weapons proliferation are argued to be invalid. Failure to complete the CRBR will have both economic and national security repercussions

  10. Fabrication of Li2TiO3 pebbles using PVA–boric acid reaction for solid breeding materials

    International Nuclear Information System (INIS)

    Highlights: • Li2TiO3 pebbles were successfully fabricated by the slurry droplet wetting method. • Boron was used as hardening agent of PVA and completely removed during sintering. • Microstructure of fabricated Li2TiO3 pebble was exceptionally homogeneous. • Suitable process conditions for high-quality Li2TiO3 pebble were summarized. - Abstract: Lithium metatitanate (Li2TiO3) is a candidate breeding material of the Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM). The breeding material is used in pebble-bed form to reduce the uncertainty of the interface thermal conductance. In this study, Li2TiO3 pebbles were successfully fabricated by the slurry droplet wetting method using the cross-linking reaction between polyvinyl alcohol (PVA) and boric acid. The effects of fabrication parameters on the shaping of Li2TiO3 green body were investigated. In addition, the basic characteristics of the sintered pebble were also evaluated. The shape of Li2TiO3 green bodies was affected by slurry viscosity, PVA content and boric acid content. The grain size and average crush load of sintered Li2TiO3 pebble were controlled by the sintering time. The boron was completely removed during the final sintering process

  11. Development of advanced blanket materials for solid breeder blanket of fusion reactor

    International Nuclear Information System (INIS)

    The design of advanced solid breeding blanket in the DEMO reactor requires the tritium breeder and neutron multiplier that can withstand the high temperature and high fluence, and the development of such as advanced blanket materials has been carried out by the cooperation activities among JAERI, universities and industries in Japan. The Li2TiO3 pebble fabricated by wet process is a reference material as a tritium breeder, but the stability on high temperature has to be improved for application to DEMO blanket. As one of such the improved materials, TiO2-doped Li2TiO3 pebbles were successfully fabricated and TiO2-doped Li2TiO3 has been studied. For the advanced neutron multiplier, the beryllides that have high melting point and good chemical stability have been studied. Some characterization of Be12Ti was conducted, and it became clear that Be12Ti had lower swelling and tritium inventory than that of beryllium metal. The pebble fabrication study for Be12Ti was also performed and Be12Ti pebbles were successfully fabricated. From these activities, the bright prospect was obtained to realize the DEMO blanket by the application of TiO2-doped Li2TiO3 and beryllides. (author)

  12. Postirradiation examination of beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    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.

  13. Formation of pebble-pile planetesimals

    CERN Document Server

    Jansson, Karl Wahlberg

    2014-01-01

    The first stage of planet formation is the accumulation of dust and ice grains into mm-cm-sized pebbles. These pebbles can clump together through the streaming instability and form gravitationally bound pebble 'clouds'. Pebbles inside such a cloud will undergo mutual collisions, dissipating energy into heat. As the cloud loses energy, it gradually contracts towards solid density. We model this process and investigate two important properties of the collapse: (i) the timescale of the collapse and (ii) the temporal evolution of the pebble size distribution. Our numerical model of the pebble cloud is zero-dimensional and treats collisions with a statistical method. We find that planetesimals with radii larger than 100 km collapse on the free-fall timescale of about 25 years. Lower-mass clouds have longer pebble collision timescales and collapse much more slowly, with collapse times of a few hundred years for 10-km-scale planetesimals and a few thousand years for 1-km-scale planetesimals. The mass of the pebble c...

  14. Fusion breeder sphere - PAC blanket design

    International Nuclear Information System (INIS)

    There is a considerable world-wide effort directed toward the production of materials for fusion reactors. Many ceramic fabrication groups are working on making lithium ceramics in a variety of forms, to be incorporated into the tritium breeding blanket which will surround the fusion reactor. Current blanket designs include ceramic in either monolithic or packed sphere bed (sphere-pac) forms. The major thrust at AECL is the production of lithium aluminate spheres to be incorporated in a sphere-pac bed. Contemporary studies on breeder blanket design offer little insight into the requirements on the sizes of the spheres. This study examined the parameters which determine the properties of pressure drop and coolant requirements. It was determined that an optimised sphere-pac bed would be composed of two diameters of spheres: 75 weight % at 3 mm and 25 weight % at 0.3 mm

  15. Conceptual design of a water cooled breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Pu, Yong; Cheng, Xiaoman [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Li, Jia; Peng, ChangHong [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); Ma, Xuebing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Chen, Lei [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China)

    2014-10-15

    Highlights: • We proposed a water cooled ceramic breeder blanket with superheated steam. • Superheated steam is generated at the first wall and the front part of breeder zone. • Superheated steam has negligible impact on neutron absorption by coolant in FW and improves TBR. • The superheated steam at higher temperature can improve thermal efficiency. - Abstract: China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by {sup 6}Li(n,α)T reaction. Li{sub 2}TiO{sub 3} pebbles and Be{sub 12}Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li{sub 2}TiO{sub 3} and Be{sub 12}Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be{sub 12}Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option

  16. Conceptual design of a water cooled breeder blanket for CFETR

    International Nuclear Information System (INIS)

    Highlights: • We proposed a water cooled ceramic breeder blanket with superheated steam. • Superheated steam is generated at the first wall and the front part of breeder zone. • Superheated steam has negligible impact on neutron absorption by coolant in FW and improves TBR. • The superheated steam at higher temperature can improve thermal efficiency. - Abstract: China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by 6Li(n,α)T reaction. Li2TiO3 pebbles and Be12Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li2TiO3 and Be12Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be12Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option, in spite of lower TBR, Pb is taken into

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

    International Nuclear Information System (INIS)

    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)

  18. Influence of chemisorption products of carbon dioxide and water vapour on radiolysis of tritium breeder

    International Nuclear Information System (INIS)

    Highlights: • Chemisorption products affect formation proceses of radiation-induced defects. • Radiolysis of chemisorption products increase amount of radiation-induced defects. • Irradiation atmosphere influence radiolysis of lithium orthosilicate pebbles. - Abstract: Lithium orthosilicate pebbles with 2.5 wt% excess of silica are the reference tritium breeding material for the European solid breeder test blanket modules. On the surface of the pebbles chemisorption products of carbon dioxide and water vapour (lithium carbonate and hydroxide) may accumulate during the fabrication process. In this study the influence of the chemisorption products on radiolysis of the pebbles was investigated. Using nanosized lithium orthosilicate powders, factors, which can influence the formation and radiolysis of the chemisorption products, were determined and described as well. The formation of radiation-induced defects and radiolysis products was studied with electron spin resonance and the method of chemical scavengers. It was found that the radiolysis of the chemisorption products on the surface of the pebbles can increase the concentration of radiation-induced defects and so could affect the tritium diffusion, retention and the released species

  19. Conceptual design of pebble drop divertor

    International Nuclear Information System (INIS)

    A pebble drop divertor concept is proposed for future fusion reactor. The marked feature of this system is the use of multi-layer pebbles that consists of a central kernel and some coating layers, as a divertor surface component. By using multi-layer pebbles, pebble drop divertor have the advantages such as steady state wall pumping with low bulk tritium retention. The performance of whole divertor system depends on the characteristics of the multi-layer pebble. Particularly the maximum heat load of the system is determined by the dimensions, the layer structure and the material of a kernel. A kernel also has an important role to determine surface temperature, which affects the wall pumping efficiency. This paper presents the numerical results of the maximum allowable heat load and the surface temperature of the divertor pebble. From the numerical estimation of thermal stress and surface temperature, it is found that the radius of divertor pebble with ceramic kernel should be 0.5 - 1 mm. (author)

  20. Conceptual design of pebble drop divertor

    International Nuclear Information System (INIS)

    A pebble drop divertor concept is proposed for future fusion reactor. The marked feature of this system is the use of multi-layer pebbles that consists of a central kernel and some coating layers, as a divertor surface component. By using multi-layer pebbles, pebble drop divertor have the advantages such as steady state wall pumping with low bulk tritium retention. The performance of whole divertor system depends on the characteristics of the multi-layer pebble. Particularly the maximum heat load of the system is determined by the dimensions, the layer structure and the material of a kernel. A kernel also has an important role to determine surface temperature, which affects the wall pumping efficiency. This paper presents the numerical results of the maximum allowable heat load and the surface temperature of the divertor pebble. From the numerical estimation of thermal stress and surface temperature, it is found that the radius of divertor pebble with ceramic kernel should be 0.5-1 mm. (author)

  1. Swiss breeder research programme

    International Nuclear Information System (INIS)

    A new initiative for a Swiss Fast Breeder Research Program has been started during 1991. This was partly the consequence of a vote in Fall 1990, when the Swiss public voted for maintaining nuclear reactors in operation, but also for a moratorium of 10 years, within which period no new reactor project should be proposed. On the other hand the Swiss government decided to keep the option 'atomic reactors' open and therefore it was essential to have programmes which guaranteed that the knowledge of reactor technology could be maintained in the industry and the relevant research organisations. There is also motivation to support a Swiss Breeder Research Program on the part of the utilities, the licensing authorities and the Paul Scherrer Institute (PSI). The utilities recognise the breeder reactor as an advanced reactor system which has to be developed further and might be a candidate, somewhere in the future, for electricity production. In so far they have great interest that a know-how base is maintained in our country, with easy access for technical questions and close attention to the development of this reactor type. The licensing authorities have a legitimate interest that an adequate knowledge of the breeder reactor type and its functions is kept at their disposal. PSI and the former EIR have had for many years a very successful basic research programme concerning breeder reactors, and were in close cooperation with EFR. The activities within this programme had to be terminated owing to limitations in personnel and financial resources. The new PSI research programme is based upon two main areas, reactor physics and reactor thermal hydraulics. In both areas relatively small but valuable basic research tasks, the results of which are of interest to the breeder community, will be carried out. The lack of support of the former Breeder Programme led to capacity problems and finally to a total termination. Therefore one of the problems which had to be solved first was

  2. Proceedings of the sixth international workshop on ceramic breeder blanket interactions

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Kenji [ed.

    1998-03-01

    This report is the Proceedings of `the Sixth International Workshop on Ceramic Breeder Blanket Interactions` which was held as a workshop on ceramic breeders under Annex II of IEA Implementing Agreement on a Programme of Research and Development on Fusion Materials, and Japan-US Workshop 97FT4-01. This workshop was held in Mito city, Japan on October 22-24, 1997. About forty experts from EU, Japan, USA, and Chile attended the workshop. The scope of the workshop included the following: (1) fabrication and characterization of ceramic breeders, (2) properties data for ceramic breeders, (3) tritium release characteristics, (4) modeling of tritium behavior, (5) irradiation effects on performance behavior, (6) blanket design and R and D requirements, (7) hydrogen behavior in materials, and (8) blanket system technology and structural materials. In the workshop, information exchange was performed for fabrication technology of ceramic breeder pebbles in EU and Japan, data of various properties of Li{sub 2}TiO{sub 3}, tritium release behavior of Li{sub 2}TiO{sub 3} and Li{sub 2}ZrO{sub 3} including tritium diffusion, modeling of tritium release from Li{sub 2}ZrO{sub 3} in ITER condition, helium release behavior from Li{sub 2}O, results of tritium release irradiation tests of Li{sub 4}SiO{sub 4} pebbles in EXOTIC-7, R and D issues for ceramic breeders for ITER and DEMO blankets, etc. The 23 of the papers are indexed individually. (J.P.N.)

  3. Shielding pebble transfer system for thermonuclear device

    International Nuclear Information System (INIS)

    In a system for supplying shielding pebbles to a vacuum vessel filled with the shielding pebbles in a gap of a double-walled structure, a supply port for the shielding pebbles is formed in a diverging shape, and a corny object is disposed at the center of the flow channel, or protrusions are formed in the vicinity of the supply port. Alternatively, a small object is disposed at the center of the flow channel of the supply port, and the small object is supported swingably and tiltably by elastic members. In addition, the upper plate of the vacuum vessel is slanted having the supply port of the shielding pebbles as a top, and a slanting angle relative to a horizontal axis is made greater than the resting angle of the shielding pebble accumulation layer. The shielding pebbles are jetted out from the supply port and spread to the peripheries, abut against the inner surface of the vacuum vessel, jump up and then accumulate. Accordingly, they can be accumulated dispersingly without being localized. An uniform accumulation layer is obtained to form a vacuum vessel having uniform and high shielding performance. (N.H.)

  4. Investigating the advantages and disadvantages of realistic approach and porous approach for closely packed pebbles in CFD simulation

    Energy Technology Data Exchange (ETDEWEB)

    Wu, C.Y. [Department of Engineering and System Science, Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Sec. 2, Kuang-Fu Rd., Hsingchu 30013, 325 Taiwan (China); Ferng, Y.M., E-mail: ymferng@ess.nthu.edu.t [Department of Engineering and System Science, Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Sec. 2, Kuang-Fu Rd., Hsingchu 30013, 325 Taiwan (China); Chieng, C.C.; Liu, C.C. [Department of Engineering and System Science, Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Sec. 2, Kuang-Fu Rd., Hsingchu 30013, 325 Taiwan (China)

    2010-05-15

    A pebble bed geometry is usually adopted for high-temperature gas-cooled reactors (HTGRs), which exhibits inherently safe performance, high conversion efficiency, and low power density design. It is important to understand the thermal-hydraulic characteristics of HTGR core for optimum design and safe operation. Therefore, this study investigates the thermal-hydraulic behaviors in a segment of pebbles predicted by the Reynolds-averaged Navier-Stokes (RANS) computational fluid dynamics (CFD) model using porous and realistic approaches for the complicated geometry. The advantages of each approach's methodology for the closely packed pebble geometry can be revealed by comparing the calculated results. In an engineering application, a CFD simulation with the porous approach for the pebble geometry can quickly and reasonably capture the averaged behaviors of the thermal-hydraulic parameters as the gas flows through the core, including the pressure drop and temperature increase. However, it is necessary to utilize the realistic approach for this complicated geometry to obtain the detailed and localized characteristics within the fluid and solid fuel regions. The present simulation results can provide useful information to help CFD researchers to determine an appropriate approach to be used when investigating the thermal-hydraulic characteristics within the reactor core of a closely packed pebble bed.

  5. Microstructure analysis of melt-based lithium orthosilicate/metatitanate pebbles

    International Nuclear Information System (INIS)

    Lithium containing ceramics, such as lithium orthosilicate (Li4SiO4) and lithium metatitanate (Li2TiO3), are being developed to be used as tritium sources for future fusion reactors. In the current design of the Helium-Cooled Pebble Bed Blanket, pebbles with a diameter of approximately 1 mm are featured in pebble beds in so called blanket modules surrounding the plasma core. A modified single drop, melt-based process has been developed for the production of the pebbles. This method has many advantages in regard to the yield and the recycling potential of the used material. The current reference material is produced with an excess of 2.5 wt. % SiO2, which results in a 2-phase material composed of roughly 90 mol % Li4SiO4 and 10 mol % Li2SiO3, lithium metasilicate. There are generally two possible methods to increase the mechanical properties of these pebbles: (a) minimize production based defects such as cracks and pores, and (b) by adding stronger phases to the ceramic. Recent work has focused on the modification of the established melt-spraying process to develop a droplet generation technique with more control over the process. Both lithium orthosilicate and metatitanate have the advantage of being low activation materials, which allows the addition of titania, TiO2, to the melt, resulting in the formation of a strengthening secondary phase of lithium metatitanate. This paper looks at the influence of the phase content on the microstructure and the resulting mechanical properties. Phase analysis and porosity measurements were performed, and in particular the microstructure and mechanical crush loads were characterized. (orig.)

  6. R and D status on Water Cooled Ceramic Breeder Blanket Technology

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio, E-mail: enoeda.mikio@jaea.go.jp; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu; Nishi, Hiroshi; Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji; Yokoyama, Kenji

    2014-10-15

    Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li{sub 2}TiO{sub 3} pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.

  7. R and D status on Water Cooled Ceramic Breeder Blanket Technology

    International Nuclear Information System (INIS)

    Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li2TiO3 pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA

  8. New reactor programs from passive to pebble bed

    International Nuclear Information System (INIS)

    The market for new nuclear power plants is small and challenged by alternative means of electric power generation. Customers and countries may vary in their requirements for a new nuclear plant; but all have a common theme of seeking a design that possesses favorable economics. This paper sets forth the economic challenges a new nuclear plant must overcome. In particular, it delineates the capital cost, construction time, and generation cost required to compete with combined cycle gas electric power generation. The U.S. power generation market is used as a point of comparison. Following this, the portfolio of BNFL/ Westinghouse plant designs are described and the methods by which they will meet the economic challenges previously delineated will be discussed. The portfolio includes the family of passive plants originated by the AP600 Design Certification process in the U.S. These plants are marked by a high degree of safety and simplicity, short construction times, and superior economics. In addition, the effort to meet European requirements for passive plants will be described. Lastly, the paper explores some advanced nuclear designs that are not yet licensed, and the hope that they hold for meeting the industry challenge ahead. (author)

  9. The modular pebble bed high temperature gas reactor

    International Nuclear Information System (INIS)

    Modular High Temperature Reactor power plants are characterized by the fact that standardized reactor units - modules -, each with a thermal power rating of 200-250 MW, can be interconnected to yield power plants in a broad power range. Provided that modular power plants are competitive, there is a variety of applications, e.g.: principal initial applications in the generation of electricity for a wide range of utility grid and plant sizes; co-generation of process steam and electricity, or district heat and electricity, for industrial or municipal consumers; and, in the long term, direct use of nuclear heat for process purposes e.g. gasifying coal, reforming methane etc. An essential condition for reasonably low capital costs is a simple design, taking into account the inherent safety features of small HTR's, e.g. the elimination of separate, redundant cooling systems for decay heat removal. Moreover, the safety concept must be simple, in order to minimize the engineering effort for the nuclear licensing procedure. Further, key reactor safety features should be convincingly demonstrated by full-scale test at an affordable cost, to provide a basis for standardized licensing of replicated reactors (the License-By-Test approach). In addition, the systems and structures within the nuclear envelope must be isolated such that the non-nuclear portion of the plant can be constructed as conventional power plant systems and structures. The Modular HTGR is designed to meet these conditions for safe, economical nuclear power

  10. New reactor programs from passive to pebble bed

    Energy Technology Data Exchange (ETDEWEB)

    Bruschi, H.J. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    2002-07-01

    The market for new nuclear power plants is small and challenged by alternative means of electric power generation. Customers and countries may vary in their requirements for a new nuclear plant; but all have a common theme of seeking a design that possesses favorable economics. This paper sets forth the economic challenges a new nuclear plant must overcome. In particular, it delineates the capital cost, construction time, and generation cost required to compete with combined cycle gas electric power generation. The U.S. power generation market is used as a point of comparison. Following this, the portfolio of BNFL/ Westinghouse plant designs are described and the methods by which they will meet the economic challenges previously delineated will be discussed. The portfolio includes the family of passive plants originated by the AP600 Design Certification process in the U.S. These plants are marked by a high degree of safety and simplicity, short construction times, and superior economics. In addition, the effort to meet European requirements for passive plants will be described. Lastly, the paper explores some advanced nuclear designs that are not yet licensed, and the hope that they hold for meeting the industry challenge ahead. (author)

  11. A Pebble-Bed Breed-and-Burn Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Greenspan, Ehud [Univ. of California, Berkeley, CA (United States)

    2016-03-31

    The primary objective of this project is to use three-dimensional fuel shuffling in order to reduce the minimum peak radiation damage of ~550 dpa present Breed-and-Burn (B&B) fast nuclear reactor cores designs (they feature 2-D fuel shuffling) call for to as close as possible to the presently accepted value of 200 dpa thereby enabling earlier commercialization of B&B reactors which could make substantial contribution to energy sustainability and economic stability without need for fuel recycling. Another objective is increasing the average discharge burnup for the same peak discharge burnup thereby (1) increasing the fuel utilization of 2-D shuffled B&B reactors and (2) reducing the reprocessing capacity required to support a given capacity of FRs that are to recycle fuel.

  12. A Pebble-Bed Breed-and-Burn Reactor

    International Nuclear Information System (INIS)

    The primary objective of this project is to use three-dimensional fuel shuffling in order to reduce the minimum peak radiation damage of ~550 dpa present Breed-and-Burn (B&B) fast nuclear reactor cores designs (they feature 2-D fuel shuffling) call for to as close as possible to the presently accepted value of 200 dpa thereby enabling earlier commercialization of B&B reactors which could make substantial contribution to energy sustainability and economic stability without need for fuel recycling. Another objective is increasing the average discharge burnup for the same peak discharge burnup thereby (1) increasing the fuel utilization of 2-D shuffled B&B reactors and (2) reducing the reprocessing capacity required to support a given capacity of FRs that are to recycle fuel.

  13. Pebble Bed Modular Reactor. Executive summary, Issue D

    International Nuclear Information System (INIS)

    The PBMR Project has been under active investigation by Eskom, as part of the Integrated Electricity Planning process, since 1993. The overall objectives of these investigations have been to establish whether such a system could form part of Eskom's expansion planning and what specific advantages it would bring over current options. This would include a technical performance and economic evaluation associated with the project. A comprehensive evaluation was performed as to the international interest existing within this field of technology, including the availability of this technology. The first phase of this investigation is now complete and the results have been compiled in a comprehensive set of technical and costing reports. A basic engineering simulator is being developed which provides engineering design support. The present results of this study show that the design has been established in enough detail to support key safety studies, confirm operating limits and estimated costing. The costing includes: - Production costs for a single module and a unit (unit = 10 x module) cost; - Operation and maintenance costing; - Fuel plant costing ? Design and development costing The studies showed that the technology required for the design has been demonstrated adequately to avoid fundamental technical risk. The increased level of inherent safety (over current designs) is fundamental to the cost reductions achieved over other nuclear designs. By demonstrating a catastrophe free design the requirements for both safety grade backup systems and an off-site emergency plan are removed

  14. Pebble bed reactor fiscal year 1980: review summary report

    International Nuclear Information System (INIS)

    Information on high-temperature reactor development is presented concerning reactor operating experience; core performance assessment; core control and shutdown; reflector and core support; maintenance and availability; safety aspects of PBR and prismatic comparison; PCRV dimensions; and fuel reprocessing cost estimate

  15. Li2TiO3 pebbles reprocessing, recovery of 6Li as Li2CO3

    International Nuclear Information System (INIS)

    A process for obtaining Li2CO3 from Li2TiO3 powder by wet chemistry was developed. This is considered useful in view of the recovery of 6Li isotope from a lithium titanate breeder burned up to its end of life in a fusion reactor. The process was optimized with respect to the chemical attack of titanate and the precipitation of carbonate from aqueous solutions to get a powder, with the chemical and morphological characteristics, suitable for its re-exploitation in the fabrication of Li2TiO3 pebbles. Reprocessing was also planned to adjust the 6Li concentration to the desired value and to obtain homogeneous distribution in the powder batch. Further development concerning reprocessing of sintered Li2TiO3 pebbles is in progress exploiting the results obtained with lithium titanate powders. (orig.)

  16. Li2TiO3 pebbles reprocessing, recovery of 6Li as Li2CO3

    International Nuclear Information System (INIS)

    A process for obtaining Li2CO3 from Li2TiO3 powder by wet chemistry was developed. This is considered useful in view of the recovery of the 6Li isotope from lithium titanate breeder burned to its end of life in a fusion reactor. The process was optimized with respect to the chemical attack of titanate and the precipitation of carbonate from aqueous solutions to get a powder with chemical and morphological characteristics suitable for its reexploitation in the fabrication of Li2TiO3 pebbles. Reprocessing was also planned to adjust the 6Li concentration to the desired value and to obtain a homogeneous distribution in the powder batch. Further development concerning reprocessing of sintered Li2TiO3 pebbles is in progress exploiting the results obtained with lithium titanate powders

  17. Neutronics and thermo-hydraulic design of supercritical-water cooled solid breeder TBM

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Jie; Wu, Yingwei, E-mail: wyw810@mail.xjtu.edu.cn; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2015-03-15

    Highlights: • A supercritical-water cooled solid breeder test blanket module (SWCB TBM) was designed. • The neutronics calculations show that the tritium breeding ratio (TBR) of SWCB TBM is 1.17. • The outlet temperature of SWCB TBM can reach as high as 500 °C. • Both thermal stress and deformation of the SWCB TBM design are within safety limits. - Abstract: In this paper, the supercritical-water cooled solid breeder test blanket module (SWCB TBM), using the supercritical water as the coolant, Li{sub 4}SiO{sub 4} lithium ceramic pebbles as a breeder, and beryllium pebbles as a neutron multiplier, was designed and analyzed for ITER. The results of neutronics, thermo-hydraulic and thermo-mechanical analysis are presented for the SWCB TBM. Neutronics calculations show that the proposed TBM has high tritium breeding ratio and power density. The tritium breeding ratio (TBR) of the proposed design is 1.17, which is greater than that of 1.15 required for tritium self-sufficiency. The thermo-hydraulic calculation proved that the TBM components can be effectively cooled to the allowable temperature with the temperature of outlet reaching 500 °C. According to thermo-mechanics calculation results, the first wall with the width of 17 mm is safe and the deformation of first wall is far below the limited value. All the results showed that the current TBM design was reasonable under the ITER normal condition.

  18. Neutronics and thermo-hydraulic design of supercritical-water cooled solid breeder TBM

    International Nuclear Information System (INIS)

    Highlights: • A supercritical-water cooled solid breeder test blanket module (SWCB TBM) was designed. • The neutronics calculations show that the tritium breeding ratio (TBR) of SWCB TBM is 1.17. • The outlet temperature of SWCB TBM can reach as high as 500 °C. • Both thermal stress and deformation of the SWCB TBM design are within safety limits. - Abstract: In this paper, the supercritical-water cooled solid breeder test blanket module (SWCB TBM), using the supercritical water as the coolant, Li4SiO4 lithium ceramic pebbles as a breeder, and beryllium pebbles as a neutron multiplier, was designed and analyzed for ITER. The results of neutronics, thermo-hydraulic and thermo-mechanical analysis are presented for the SWCB TBM. Neutronics calculations show that the proposed TBM has high tritium breeding ratio and power density. The tritium breeding ratio (TBR) of the proposed design is 1.17, which is greater than that of 1.15 required for tritium self-sufficiency. The thermo-hydraulic calculation proved that the TBM components can be effectively cooled to the allowable temperature with the temperature of outlet reaching 500 °C. According to thermo-mechanics calculation results, the first wall with the width of 17 mm is safe and the deformation of first wall is far below the limited value. All the results showed that the current TBM design was reasonable under the ITER normal condition

  19. Lithium titanate pebbles reprocessing by wet chemistry

    International Nuclear Information System (INIS)

    An original dissolution method for irradiated Li2TiO3 in aqueous H2O2 was developed. One could easily obtain fine Li2TiO3 powders from the solution through drying and calcination. Li2TiO3 pebbles (size ∼0.6 mm, above 90% TD) were obtained from the 'reprocessed' powders. These solutions were also suitable for the formation of a sol emulsion in 2-ethyl-hexanol-1, from which gelled microspheres of lithium titanate could be obtained. Locally prepared Li2TiO3 reprocessed and supplied pebble batches were tested for tritium release by temperature programmed desorption (TPD) methods in He + 0.1%H2 (R-gas) after their short irradiations in a thermal neutron flux. The relative TPD data were compared. A qualitative correlation was developed between peak characteristics and pebble microstructure

  20. Fuel feeding with pebbles, installation and experience

    International Nuclear Information System (INIS)

    The AVR reactor is a graphite moderated high temperature reactor cooled with helium (at 10 bar). It was the first reactor which operated with spherical fuel elements. The fuel elements have a diameter of 60 mm. They consists of graphite in which the fuel is embedded as coated particles, and weigh about 200 g. The core is cylindrical, with a diameter of 3 m and 3 m high. About 100,000 pebbles are accommodated in it. The fuelling equipment differs considerably from the fuelling and fuel removal machines of previously operated reactors with rod-shaped or block-shaped fuel elements, because of the spherical fuel elements. The tasks of the fuelling equipment are the addition, turning over and removal of fuel elements and test pebbles. The burnup also has to be measured and all pebbles have to be counted and recorded. One can talk of a completely new development, as there were no tested components for all these tasks. (orig.)

  1. Can the breeder go commercial

    International Nuclear Information System (INIS)

    Contrary to some beliefs in the electric utility industry that ERDA is committed to developing a commercial breeder economy, it is pointed out that ERDA isn't even willing to pay the total cost of the R and D program--and unless there is a major commitment from the private sector (the electric utility industry, in particular) the breeder program will die. The schedule as of Fall 1976 called for: (1) Fast Flux Test Facility (scheduled to go critical in 1979, operate in 1980); (2) Clinch River Breeder Reactor Project (CRBRP) (1/3 commercial size plant hopefully operating by 1983); (3) Prototype Large Breeder Reactor (planned construction starting in 1981, operating in 1988); and (4) Commercial Breeder Reactor (CBR-1 design work to start in 1983, construction in 1986, and operation in 1993). The $257 million the utility industry has pledged to the CRBRP was just for openers. The $2 billion follow-on breeder project being designed calls for massive capital input from a utility (or utility consortium)--and if that is not forthcoming, then in the words of an ERDA official, ''we'll have to reassess the whole breeder program.''

  2. International strategies for breeder development

    International Nuclear Information System (INIS)

    This paper studies the perspectives of breeder reactors development. The near term context has led some experts to the conclusion that breeder reactor technology is too far ahead of its time. Some have compared breeders to the supersonic airplane, Concorde: good technical performance but failure in its economic dimensions. In this paper, the author points out the major shortcomings of such an assessment which may be valid in the short time. However, with a short-term market-dominated perspective that uses an 8% discount rate, one can neglect every thing that is going to happen in 50 years. 6 refs., 11 figs

  3. Fabrication of Li4SiO4 pebbles by wet method with modified powders synthesized via sol–gel process

    International Nuclear Information System (INIS)

    Li4SiO4 pebbles have been recognized as attractive tritium breeder materials in the fusion reactor blanket of international thermonuclear experimental reactor (ITER). In this work, we present a facile method to prepare Li4SiO4 pebbles of high density and sphericity by using a directive wet method with the Li4SiO4 powders synthesized via sol–gel process. The Li4SiO4 powders were prepared with two-step calcinating method, followed by a ball-milling process. Thermal and phase analysis, morphologies and sintering behaviors observations of the pebbles were carried out systematically. Experimental results show that the pure phase powders with white color that prepared by using two-step calcinating method is different from the powders prepared by the traditional direct calcinating method. The subsequent ball milling process proves to be effective to improve the relative density of the sintered body. When sintered at the temperature as low as 850 °C for 4 h, the favorable Li4SiO4 pebbles with uniform size (∼1 mm), good sphericity (1.02), and high density (above 90% T.D.) were fabricated by using a directive wet method. The as-fabricated pebbles hold good potential as tritium breeding materials for blankets

  4. Formation and accumulation of radiation-induced defects and radiolysis products in modified lithium orthosilicate pebbles with additions of titanium dioxide

    Science.gov (United States)

    Zarins, Arturs; Valtenbergs, Oskars; Kizane, Gunta; Supe, Arnis; Knitter, Regina; Kolb, Matthias H. H.; Leys, Oliver; Baumane, Larisa; Conka, Davis

    2016-03-01

    Lithium orthosilicate (Li4SiO4) pebbles with 2.5 wt.% excess of silicon dioxide (SiO2) are the European Union's designated reference tritium breeding ceramics for the Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM). However, the latest irradiation experiments showed that the reference Li4SiO4 pebbles may crack and form fragments under operation conditions as expected in the HCPB TBM. Therefore, it has been suggested to change the chemical composition of the reference Li4SiO4 pebbles and to add titanium dioxide (TiO2), to obtain lithium metatitanate (Li2TiO3) as a second phase. The aim of this research was to investigate the formation and accumulation of radiation-induced defects (RD) and radiolysis products (RP) in the modified Li4SiO4 pebbles with different contents of TiO2 for the first time, in order to estimate and compare radiation stability. The reference and the modified Li4SiO4 pebbles were irradiated with accelerated electrons (E = 5 MeV) up to 5000 MGy absorbed dose at 300-990 K in a dry argon atmosphere. By using electron spin resonance (ESR) spectroscopy it was determined that in the modified Li4SiO4 pebbles, several paramagnetic RD and RP are formed and accumulated, like, E' centres (SiO33-/TiO33-), HC2 centres (SiO43-/TiO3-) etc. On the basis of the obtained results, it is concluded that the modified Li4SiO4 pebbles with TiO2 additions have comparable radiation stability with the reference pebbles.

  5. Tritium-assisted fusion breeders

    International Nuclear Information System (INIS)

    This report undertakes a preliminary assessment of the prospects of tritium-assisted D-D fuel cycle fusion breeders. Two well documented fusion power reactor designs - the STARFIRE (D-T fuel cycle) and the WILDCAT (Cat-D fuel cycle) tokamaks - are converted into fusion breeders by replacing the fusion electric blankets with 233U producing fission suppressed blankets; changing the Cat-D fuel cycle mode of operation by one of the several tritium-assisted D-D-based modes of operation considered; adjusting the reactor power level; and modifying the resulting plant cost to account for the design changes. Three sources of tritium are considered for assisting the D-D fuel cycle: tritium produced in the blankets from lithium or from 3He and tritium produced in the client fission reactors. The D-D-based fusion breeders using tritium assistance are found to be the most promising economically, especially the Tritium Catalyzed Deuterium mode of operation in which the 3He exhausted from the plasma is converted, by neutron capture in the blanket, into tritium which is in turn fed back to the plasma. The number of fission reactors of equal thermal power supported by Tritium Catalyzed Deuterium fusion breeders is about 50% higher than that of D-T fusion breeders, and the profitability is found to be slightly lower than that of the D-T fusion breeders

  6. Gas bubble network formation in irradiated beryllium pebbles monitored by X-Ray micro-tomography

    International Nuclear Information System (INIS)

    Full text of publication follows: The efficient and safe operation of helium cooled ceramic breeder blankets requires among others an efficient tritium release during operation at blanket relevant temperatures. In the past out-of-pile thermal desorption studies on low temperature neutron irradiated beryllium have shown that tritium and helium release peaks occur together. This phenomenon can be interpreted in terms of growth and coalescence of helium bubbles and tritium that either is trapped inside the helium bubbles in form of T2 molecules or in their strain field. With increasing temperature the bubble density and size at grain interfaces increase together with the probability of interconnected porosities and channel formation to the outer surface, leading to simultaneous helium and tritium release peaks in TDS. For a reliable prediction of gas release up to end-of-life conditions at blanket relevant temperatures, knowledge of the dynamics of bubble growth and coalescence as well as the 3D distribution of bubble network formation is indispensable. Such data could also be used to experimentally validate any future model predictions of tritium and helium release rates. A high resolution computer aided micro-tomography (CMT) setup has been developed at the European Synchrotron Radiation Facility which allowed reconstructing 3-D images of beryllium pebbles without damaging them. By postprocessing the data a 3D rendering of inner surfaces and of interconnected channel networks can be obtained, thus allowing the identification of open porosities in neutron irradiated and tempered beryllium pebbles. In our case Beryllium pebbles of 2 mm diameter had been neutron irradiated in the 'Beryllium' experiment at 770 K with 1.24 x 1025 nxm-2 resulting in 480 appm He and 12 appm Tritium. After annealing at 1500 K CMT was performed on the pebbles with 4.9 and 1.4 μm voxel resolution, respectively, followed by morphological and topological post-processing of the reconstructed

  7. Pebble Accretion and the Diversity of Planetary Systems

    CERN Document Server

    Chambers, J E

    2016-01-01

    I examine the standard model of planet formation, including pebble accretion, using numerical simulations. Planetary embryos large enough to become giant planets do not form beyond the ice line within a typical disk lifetime unless icy pebbles stick at higher speeds than in experiments using rocky pebbles. Systems like the Solar System (small inner planets, giant outer planets) can form if (i) icy pebbles are stickier than rocky pebbles, and (ii) the planetesimal formation efficiency increases with pebble size, which prevents the formation of massive terrestrial planets. Growth beyond the ice line is dominated by pebble accretion. Most growth occurs early, when the surface density of pebbles is high due to inward drift of pebbles from the outer disk. Growth is much slower after the outer disk is depleted. The outcome is sensitive to the disk radius and turbulence level, which control the lifetime and maximum size of pebbles. The outcome is sensitive to the size of the largest planetesimals since there is a th...

  8. International cooperation on breeder reactors

    International Nuclear Information System (INIS)

    In March 1977, as the result of discussions which began in the fall of 1976, the Rockefeller Foundation requested International Energy Associates Limited (IEAL) to undertake a study of the role of international cooperation in the development and application of the breeder reactor. While there had been considerable international exchange in the development of breeder technology, the existence of at least seven major national breeder development programs raised a prima facie issue of the adequacy of international cooperation. The final product of the study was to be the identification of options for international cooperation which merited further consideration and which might become the subject of subsequent, more detailed analysis. During the course of the study, modifications in U.S. breeder policy led to an expansion of the analysis to embrace the pros and cons of the major breeder-related policy issues, as well as the respective views of national governments on those issues. The resulting examination of views and patterns of international collaboration emphasizes what was implicit from the outset: Options for international cooperation cannot be fashioned independently of national objectives, policies and programs. Moreover, while similarity of views can stimulate cooperation, this cannot of itself provide compelling justification for cooperative undertakings. Such undertakings are influenced by an array of other national factors, including technological development, industrial infrastructure, economic strength, existing international ties, and historic experience

  9. Sensitivity and Uncertainty Analyses of the Tritium Production in the HCPB Breeder Blanket Mock-up Experiment

    International Nuclear Information System (INIS)

    Dedicated computational methods, tools and data have been recently developed in the framework of the European Fusion Technology Programme to enable sensitivity and uncertainty analyses of fusion neutronics experiments. severely limited due to these two requirements. (author)er productgeneration and the associated uncertainties against the experimental data provided in the neutronics experiment at the Frascati Neutron Generator on a mock-up of the HCPB (Helium-Cooled Pebble Bed) breeder test blanket. This work is devoted to the computational analyses of this experiment comprising the following steps: (i) Calculation of the Tritium production rates (TPR) in the Li2CO3 pellets using a detailed 3D model of the experimental set-up; the Monte Carlo code MCNP and the discrete ordinates code TORT were applied for these calculations with EFF-3 and FENDL-2.0/2.1 nuclear data. (ii) Sensitivity calculations for the Li2CO3 pellets stacks to assess the sensitivity of the Tritium production to the reactions cross-sections of the involved nuclides Be, 6,7Li, C and O; the calculations were performed with the MCSEN Monte Carlo code using the track length estimator and, in parallel, with the deterministic SUSD3D code using neutron fluxes calculated by TORT in forward and adjoint mode. (iii) Calculations of the data related uncertainties of the TPR using co-variance data from EFF (9Be, 6Li, 12C), FENDL-2 (7Li) and JENDL-3.3 (16O); both probabilistic (MCNP/MCSEN) and deterministic (TORT/SUSD3D) approaches were applied. (iv) Assessment of the total uncertainties for the TPR including uncertainties of the measurements, the nuclear data and the calculations. The data related uncertainties of the calculated Tritium generation are in the order of 4 - 5 % (2 sigma). The main uncertainties are due to the Be cross-section data. The total uncertainties of the predicted TPR including data uncertainties, statistical uncertainties of the Monte Carlo calculation and the experimental uncertainties

  10. Preliminary Neutronics Analysis Of Fuel Pebble With Thorium Fuel Cycle

    International Nuclear Information System (INIS)

    A new fuel pebble was designed based on Thorium fuel cycle. 231Pa has been added into fuel pebble for obtaining the minimum reactivity swing. The results show that the new designed pebble fuel with 7.0 % 233U enrichment adding 3.2% 231Pa, the keff is to be controlled up to 65 GWd/t; the other design with 8.0 % 233U enrichment requires 3.9% 231Pa, the keff therefore is remain up to 80 GWd/t. About 95% of loaded 231Pa in fuel pebble is depleted after 120 GWd/t. The results imply that it is optimistic to design the fuel pebble with 233U, 231Pa and 232Th; but some effects such as fuel temperature effect, distribution of TRISO particle in pebble fuel, etc. are required to investigate. (author)

  11. Fusion Breeder Program interim report

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.; Lee, J.D.; Neef, W.

    1982-06-11

    This interim report for the FY82 Fusion Breeder Program covers work performed during the scoping phase of the study, December, 1981-February 1982. The goals for the FY82 study are the identification and development of a reference blanket concept using the fission suppression concept and the definition of a development plan to further the fusion breeder application. The context of the study is the tandem mirror reactor, but emphasis is placed upon blanket engineering. A tokamak driver and blanket concept will be selected and studied in more detail during FY83.

  12. Fusion Breeder Program interim report

    International Nuclear Information System (INIS)

    This interim report for the FY82 Fusion Breeder Program covers work performed during the scoping phase of the study, December, 1981-February 1982. The goals for the FY82 study are the identification and development of a reference blanket concept using the fission suppression concept and the definition of a development plan to further the fusion breeder application. The context of the study is the tandem mirror reactor, but emphasis is placed upon blanket engineering. A tokamak driver and blanket concept will be selected and studied in more detail during FY83

  13. Hydrodynamic and heat transfer mathematical model for the pebble-bod high temperature reactor core

    International Nuclear Information System (INIS)

    Mathematical model for the pebble-bed reactor core in a two-dimensional approximation under the coolant natural circulation conditions is considered. The calculation region represents a reactor core with a lateral reflector and unloading channel through which counercurrent gas flow enters. As basic physical model is the porous body model. Thermal hydraulic reactor core calculation for the normal operation mode on the base of the suggested mathematical model is in perfect agreement with the results of the experiment and calculations obtained by other methods. The results of calculation under the coolant natural circulation regime are in qualitative agreement with the analytical estimations obtained by the perturbation method

  14. Pebble Delivery for Inside-Out Planet Formation

    CERN Document Server

    Hu, Xiao; Chatterjee, Sourav

    2014-01-01

    Inside-Out Planet Formation (IOPF; Chatterjee & Tan 2014, hereafter CT14) is a scenario for sequential in situ planet formation at the pressure traps of retreating dead zone inner boundaries (DZIBs) motivated to explain the many systems with tightly packed inner planets (STIPs) discovered by Kepler. The scenario involves build-up of a pebble-dominated protoplanetary ring, supplied by radial drift of pebbles from the outer disk. It may also involve further build-up of planetary masses to gap-opening scales via continued pebble accretion. Here we study radial drift & growth of pebbles delivered to the DZIB in fiducial IOPF disk models.

  15. Tritium analyses of COBRA-1A2 beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Baldwin, D.L. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    Selected tritium measurements have been completed for the COBRA-1A2 experiment C03 and D03 beryllium pebbles. The completed results, shown in Tables 1, 2, and 3, include the tritium assay results for the 1-mm and 3-mm C03 pebbles, and the 1-mm D03 pebbles, stepped anneal test results for both types of 1-mm pebbles, and the residual analyses for the stepped-anneal specimens. All results have been reported with date-of-count and are not corrected for decay. Stepped-anneal tritium release response is provided in addenda.

  16. Thermal cycle test of elemental mockups of ITER breeding blanket

    International Nuclear Information System (INIS)

    Thermal cycle tests for mockups of breeder pebble beds of ITER breeding blanket have been carried out to investigate their thermo-mechanical behavior with the interaction between a pebble bed and a breeder rod containing the breeder pebbles. The mockups have been designed to demonstrate a part of the Breeder Inside Tube (BIT)' structure of ITER breeding blanket. Candidate material pebbles of Li2TiO3 was applied as breeder specimen, and Al pebbles were applied for simulating the neutron multiplier of Be pebbles. These pebbles have been packed in test tubes by using a vibration machine. Tested configurations were single layer mockups with Li2TiO3 single diameter packing and binary packing beds, and double layer mockups with Li2TiO3/Al single diameter packing and binary packing beds. In order to clarify the deformation performance of breeder tube, two different thickness of the breeder rod were also tested: one for nominal condition and another for acceleration test. Pebble bed of Li2TiO3 is heated with an electric heater, which is equipped at the center of the breeder rod, simulating the temperature profile by volumetric heating of breeder pebbles. The outside of a breeder rod in a single layer mockup and the outside of the outer tube in case of double layer mockup is cooled by water. Temperature of the breeder beds has been controlled by a power input of the heater. After the thermal cycle tests, the internal dimensions and local packing fraction of mockups have been examined by using an X-ray CT device. As the result, no significant change of packing fraction was observed after five thermal cycles with maximum heater temperature of 600degC. Any bulging of the breeder rod or any cracking of the pebble has not been observed. A soundness of the typical structure and breeder pebble bed of ITER breeding blanket against thermal cycles was confirmed. (author)

  17. EXOTIC-7: Irradiation of ceramic breeder materials to high lithium burnup

    International Nuclear Information System (INIS)

    The EXOTIC-7 irradiation experiment in the High Flux Reactor (HFR) at Petten has been completed. Its aim has been to investigate the effects of high lithium-burnup on the mechanical stability and tritium release characterisitcs of candidate ceramic breeder materials, originating from the Fusion Programmes of CEA, FZK, ENEA, AECL and ECN. The tested ceramic breeder materials were pellets of Li2ZrO3, LiAlO2 and Li8ZrO6 and pebbles of Li4SiO4 and Li2ZrO3, with a variety of characteristics, like grain size and porostiy. The test matrix provided the simultaneous irradiation of eight independent capsules with on-line tritium monitoring. Two capsules containd a mixture of Li4SiO4 and beryllium pebbles. The experimental design, sample loading and main riiadiation parameters are described. Some PIE results and analysis of in-situ tritium release behaviour are presented. (orig.)

  18. Recent progress in safety assessments of Japanese water cooled solid breeder test blanket module

    International Nuclear Information System (INIS)

    Water Cooled Solid Breeder Test Blanket Module (WCSB TBM) is being designed by JAEA for the primary candidate TBM of Japan, and the safety evaluation of WCSB TBM has been performed. This reports presents summary of safety evaluation activities of the Japanese WCSB TBM, including nuclear analysis, source of RI, waste evaluation, occupational radiolysis exposure (ORE), failure mode effect analysis (FMEA) and postulated initiating event (PIE). For the purpose of basic evaluation of source terms on nuclear heating and radioactivity generation, two-dimensional nuclear analysis has been carried out. By the nuclear analysis, distributions of neutron flux, tritium breeding ratio (TBR), nuclear heat, decay heat and induced activity are calculated. Tritium production is calculated by the nuclear analysis by integrating distributions of TBR values, as about 0.2 g-T/FPD. With respect to the radioactive waste, the induced activity of the irradiated TBM is estimated. For the purpose of occupational radiolysis exposure (ORE), RI inventory is estimated. Tritium inventory in pebble bed of TBM is about 3 x 1012 Bq, and tritium in purge gas is about 3 x 1011 Bq. FMEA has been carried out to identify the PIEs that need safety evaluation. PIEs are summarized into three groups, i.e., heating, pressurization and release of RI. PIEs of local heating are converged without any special cares. With respect to heating of whole module, two PIEs are selected as the most severe events, i.e., loss of cooling of TBM during plasma operation and ingress of coolant into TBM during plasma operation. With respect to PIEs about pressurization, the PIEs of pressurization of the compartment nearby the pipes of cooling system are evaluated, because rupture of the pipes result pressurization of such compartments, i.e., box structure of TBM, purge gas loop, TRS, VV, port cell and TCWS vault. Box structure of TBM is designed to withstand the maximum pressure of the cooling system. At other compartments

  19. Fast breeder reactor research

    International Nuclear Information System (INIS)

    , Italy, in April or May 1977. Recognizing the importance of international co-ope ration within the framework of IWGFR for preparing surveys, proposals and recommendations concerning sodium cooled fast breeder reactors, the Working Group prepared a number of joint documents with the help of experts from the participating countries, discussed them at the Eighth Annual Meeting and made recommendations on the preparation of subsequent joint documents. (author)

  20. Results of AVR fuel pebble irradiation at increased temperature and burn-up in the HFR Petten

    International Nuclear Information System (INIS)

    The irradiation experiment HFR-EU1bis was performed by the European Commission's Joint Research Centre-Institute for Energy (JRC-IE) in the HFR Petten to test five spherical High Temperature Reactor (HTR) fuel pebbles of former German production with TRISO coated particles for their potential for very high temperature performance and high burn-up. The irradiation started on 9 September 2004 and was terminated on 18 October 2005 after 10 reactor cycles totaling 249 efpd and a maximum burn-up of 11.07% FIMA. The objective of the HFR-EU1bis test was to irradiate five HTR fuel pebbles at conditions beyond the characteristics of current HTR reactor designs with pebble bed cores, e.g. HTR-Modul, HTR-10 and PMBR. This should demonstrate that pebble bed HTRs are capable of enhanced performance in terms of sustainability (further increased power conversion efficiency, better use of fuel) and thus reduced waste production. The central temperature of all pebbles was kept as closely as possible at 1250 deg. C and held constant during the entire irradiation, with the exception of HFR downtime and power transients. This is the expected maximum central fuel temperature of a pebble bed VHTR with a coolant outlet temperature of 1000 deg. C. HFR-EU1bis should demonstrate the feasibility of low coated particle failure fractions under normal operating conditions and more specifically: - increased central fuel temperature of 1250 deg. C compared to 1000-1200 deg. C in earlier irradiation tests; - irradiation to a burn-up close to 16% FIMA, which is double the license limit of the HTR-Modul; due to a neutronics data processing error, the experiment was prematurely terminated at 11.07% FIMA maximum so that this objective was not fully achieved; - confirmation of low coated particle failure fractions due to temperature, burn-up and neutron fluence. This paper provides the irradiation history of the experiment including data on fission gas release. Post-irradiation examinations at NRG

  1. Numerical-experimental analyses by Hot-Wire method of an alumina cylinder for future studies on thermal conductivity of the fusion breeder materials

    International Nuclear Information System (INIS)

    The determination of the thermal conductivity of breeder materials is one of the main goal in order to find the best candidate material for the fusion reactor technology. Experimental tests have been and will be carried out with a dedicated experimental devices, built at the Department of Civil and Industrial Engineering of the University of Pisa. The methodological approach used in doing that is characterized by two main phases strictly interrelated each other: the first one focused on the experimental evaluation of thermal conductivity of a ceramic material, by means of hot wire method, to be subsequently used in the second phase, based on the test rig method, to determine the thermal conductivity of pebble bed material. To the purpose, two different experimental devices have been designed and built. This paper deals with the first phase of the methodology. In this framework, the equipment set up and built to perform Hot wire tests, the ceramic material (a cylinder of alumina), the experimental procedure and the measured results obtained varying the temperature, are presented and discussed. The experimental campaign has been lead from 50°C up to 400°C. The thermal conductivity of the ceramic material at different bulk temperatures has been obtained in stationary conditions (detected on the basis of the temperature values measured during the experiment). Numerical analyses have been also performed by means of FEM code Ansys©. The numerical results were in quite good agreement with the experimental one, confirming also the reliability of code in reproducing heat transfer phenomena

  2. Tritium breeding mock-up experiments containing lithium titanate ceramic pebbles and lead irradiated with DT neutrons

    International Nuclear Information System (INIS)

    Highlights: • Breeding benchmark experiment on LLCB TBM in ITER was performed. • Nuclear responses measured are TPR and reaction rate of 115In(n, n′)115mIn reaction. • Measured responses are compared with calculations by MCNP and FENDL 2.1 library. • TPR measurements agree with calculations in the estimated error bar. • Measured 115In(n, n′)115mIn reaction rates are underestimated by the calculations. - Abstract: Experiments were conducted with breeding blanket mock-up consisting of two layers of breeder material lithium titanate pebbles and three layers of pure lead as neutron multiplier. The radial dimensions of breeder, neutron multiplier and structural material layers are similar to the current design of the Indian Lead–Lithium cooled Ceramic Breeder (LLCB) blanket. The mock-up assembly was irradiated with 14 MeV neutrons from DT neutron generator. The local tritium production rates (TPR) from 6Li and 7Li in breeder layers were measured with the help of two different compositions of Li isotopes (60.69% 6Li and 7.54% 6Li) in Li2CO3. Tritium production in the multiplication layers were also measured with above mentioned two types of pellets to compare the experimental tritium production with calculations. TPR from 6Li at one location in the breeder layer was also measured by direct online measurement of tritons from 6Li(n, t)4He reaction using silicon surface barrier detector and 6Li to triton converter. Additional verification of neutron spectra (En > 0.35 MeV) in the mock-up zones were obtained by measuring 115In(n, n′)115mIn reaction rate and comparing it with calculated values in all five layers of mock-up. All the measured nuclear responses were compared with transport calculations using code MCNP with FENDL2.1 and FENDL3.0 cross-section libraries. The average C/E ratio for tritium production in enriched Li2CO3 pellets was 1.11 in first breeder zone and 1.09 in second breeder zone with uncertainty 8.3% at 1σ level. The experimental details

  3. Tritium breeding mock-up experiments containing lithium titanate ceramic pebbles and lead irradiated with DT neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Jakhar, Shrichand; Abhangi, M.; Tiwari, S. [Institute for Plasma Research, Bhat, Gandhinagar 382 428 (India); Makwana, R. [Department of Physics, MS University, Vadodara (India); Chaudhari, V.; Swami, H.L.; Danani, C.; Rao, C.V.S.; Basu, T.K. [Institute for Plasma Research, Bhat, Gandhinagar 382 428 (India); Mandal, D.; Bhade, Sonali; Kolekar, R.V.; Reddy, P.J. [Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Bhattacharyay, R.; Chaudhuri, P. [Institute for Plasma Research, Bhat, Gandhinagar 382 428 (India)

    2015-06-15

    Highlights: • Breeding benchmark experiment on LLCB TBM in ITER was performed. • Nuclear responses measured are TPR and reaction rate of {sup 115}In(n, n′){sup 115m}In reaction. • Measured responses are compared with calculations by MCNP and FENDL 2.1 library. • TPR measurements agree with calculations in the estimated error bar. • Measured {sup 115}In(n, n′){sup 115m}In reaction rates are underestimated by the calculations. - Abstract: Experiments were conducted with breeding blanket mock-up consisting of two layers of breeder material lithium titanate pebbles and three layers of pure lead as neutron multiplier. The radial dimensions of breeder, neutron multiplier and structural material layers are similar to the current design of the Indian Lead–Lithium cooled Ceramic Breeder (LLCB) blanket. The mock-up assembly was irradiated with 14 MeV neutrons from DT neutron generator. The local tritium production rates (TPR) from {sup 6}Li and {sup 7}Li in breeder layers were measured with the help of two different compositions of Li isotopes (60.69% {sup 6}Li and 7.54% {sup 6}Li) in Li{sub 2}CO{sub 3}. Tritium production in the multiplication layers were also measured with above mentioned two types of pellets to compare the experimental tritium production with calculations. TPR from {sup 6}Li at one location in the breeder layer was also measured by direct online measurement of tritons from {sup 6}Li(n, t){sup 4}He reaction using silicon surface barrier detector and {sup 6}Li to triton converter. Additional verification of neutron spectra (E{sub n} > 0.35 MeV) in the mock-up zones were obtained by measuring {sup 115}In(n, n′){sup 115m}In reaction rate and comparing it with calculated values in all five layers of mock-up. All the measured nuclear responses were compared with transport calculations using code MCNP with FENDL2.1 and FENDL3.0 cross-section libraries. The average C/E ratio for tritium production in enriched Li{sub 2}CO{sub 3} pellets was 1

  4. Status and prospects of thermal breeders

    International Nuclear Information System (INIS)

    The main objective of this cooperative study and of this report is to evaluate the extent to which thermal breeders might complement or serve as an alternative to fast breeders in solving the long-term nuclear fuel supply problem. A secondary objective is to consider in a general way issues such as proliferation, safety, environmental impacts, economics, power plant availability, and fuel cycle versatility to determine whether thermal breeder reactors offer advantages or disadvantages with respect to such issues

  5. Fusion breeder neutronics. Final report

    International Nuclear Information System (INIS)

    Research efforts in fusion breeder neutronics have been focused on two tasks that are strongly related. Efforts in Task 1 concentrate on examining the required conditions to sustain fuel self-sufficiency in fusion reactors operated on a D-T fuel cycle. In this respect, in-depth and detailed engineering analyses have been performed on various blanket and reactor concepts to verify the potential of each blanket concept to exhibit a tritium breeding ratio (TBR) in excess of unity by a margin that compensates for losses, radioactive decay and other inventory requirements. Efforts in Task 2 concentrate on evaluating the overall uncertainties (both experimental and analytical) associated with the TBR

  6. Building massive compact planetesimal disks from the accretion of pebbles

    CERN Document Server

    Moriarty, John

    2015-01-01

    We present a model in which planetesimal disks are built from the combination of planetesimal formation and accretion of radially drifting pebbles onto existing planetesimals. In this model, the rate of accretion of pebbles onto planetesimals quickly outpaces the rate of direct planetesimal formation in the inner disk. This allows for the formation of a high mass inner disk without the need for enhanced planetesimal formation or a massive protoplanetary disk. Our proposed mechanism for planetesimal disk growth does not require any special conditions to operate. Consequently, we expect that high mass planetesimal disks form naturally in nearly all systems. The extent of this growth is controlled by the total mass in pebbles that drifts through the inner disk. Anything that reduces the rate or duration of pebble delivery will correspondingly reduce the final mass of the planetesimal disk. Therefore, we expect that low mass stars (with less massive protoplanetary disks), low metallicity stars and stars with gian...

  7. On the growth of pebble-accreting planetesimals

    CERN Document Server

    Visser, Rico G

    2015-01-01

    Pebble accretion is a new mechanism to quickly grow the cores of planets. In pebble accretion, gravity and gas drag conspire to yield large collisional cross sections for small particles in protoplanetary disks. However, before pebble accretion commences, aerodynamical deflection may act to prevent planetesimals from becoming large, because particles tend to follow gas streamlines. We derive the planetesimal radius where pebble accretion is initiated and determine the growth timescales of planetesimals by sweepup of small particles. We obtain the collision efficiency factor as the ratio of the numerically-obtained collisional cross section to the planetesimal surface area, from which we obtain the growth timescales. Integrations are conducted in the potential flow limit (steady, inviscid) and in the Stokes flow regime (steady, viscid). Only particles of stopping time $t_s \\ll t_X$ where $t_X\\approx10^3$ s experience aerodynamic deflection. Even in that case, the planetesimal's gravity always ensures positive ...

  8. Reconstructing the transport history of pebbles on Mars

    OpenAIRE

    Szabó, Tímea; Domokos, Gábor; Grotzinger, John P.; Jerolmack, Douglas J.

    2015-01-01

    The discovery of remarkably rounded pebbles by the rover Curiosity, within an exhumed alluvial fan complex in Gale Crater, presents some of the most compelling evidence yet for sustained fluvial activity on Mars. While rounding is known to result from abrasion by inter-particle collisions, geologic interpretations of sediment shape have been qualitative. Here we show how quantitative information on the transport distance of river pebbles can be extracted from their shape alone, using a combin...

  9. Steam chemical reactivity of Be pebbles and Be powder

    International Nuclear Information System (INIS)

    This paper reports the results of chemical reactivity experiments for Be pebbles (2 and 0.2 mm diameter) and Be powder (14-31 μm diameter) exposed to steam at elevated temperatures, 350-900 deg. C for pebbles and 400-500 deg. C for powders. We measured BET specific surface areas of 0.12 m2/g for 2 mm pebbles, 0.24 m2/g for 0.2 mm pebbles and 0.66-1.21 m2/g for Be powder samples. These experiments showed a complex reactivity behavior for the material, dependent primarily on the test temperature. Average H2 generation rates for powder samples, based on measured BET surface areas, were in good agreement with previous measurements for fully dense consolidated powder metallurgy (CPM)-Be. Rates for the Be pebbles, based on measured BET surface areas, were systematically lower than the CPM-Be rates, possibly because of different surface and bulk features for the pebbles, especially surface layer impurities, that contribute to the measured BET surface area and influence the oxidation process at the material surface

  10. Global depletion analysis of Korean helium cooled solid breeder TBM model for demo fusion reactor

    International Nuclear Information System (INIS)

    The Korean HCSB (helium cooled solid breeder) TBM (test blanket module) is proposed with its specific compositions of lithium ceramic, beryllium and graphite in pebble form. In the Korean HCSB TBM, the amount of beryllium is reduced and the reduction is replaced by graphite for a neutron reflector, while tritium breeding ratio (TBR) remains almost unchanged with relatively low Li6 enrichment of ∼40%. However, the previous Korean HCSB was designed based on the LOCAL assumption, in which the surroundings are assumed by the reflective boundary condition. In this research, we establish a simple GLOBAL neutronics model based on demo fusion reactor and perform neutronics analyses including depletion (transmutation) calculation during 100 EFPDs (effective full power days) using the modified MONTEBURNS code.

  11. Preliminary results of the HFR-EU1 fuel irradiation of INET and AVR pebbles in the HFR Petten - HTR2008-58049

    International Nuclear Information System (INIS)

    The irradiation experiment HFR-EU1 in the HFR Petten is currently being conducted by the European Commission's Joint Research Centre - Inst. for Energy (JRC-IE). The irradiation targets are 5 spherical High Temperature Reactor (HTR) fuel pebbles. 2 of INET production and 3 of former German production. Both types are made of TRISO coated particles and are tested for their potential for very high temperature performance and high burn-up. The irradiation started on 29 September 2006 and, by 24 February 2008, had accumulated 12 reactor cycles totaling 332.8 efpd and a calculated maximum burn-up of 8.9% FIMA (INET) and 11.2% FIMA (AVR). The objective of the HFR-EU1 test is to irradiate 5 HTR fuel pebbles at conditions beyond the characteristics of current HTR reactor designs with pebble bed cores, e.g. HTR-Modul, HTR-10 and PMBR. This should demonstrate that pebble bed HTRs are capable of enhanced performance in terms of sustainability (further increased power conversion efficiency, improved fuel use) and thus reduced waste production. The surface temperature of all pebbles was held constant during the irradiation, with the exception of HFR downtime and power transients. HFR-EU1 should demonstrate the feasibility of low coated particle failure fractions under normal operating conditions and more specifically: - high fuel surface temperature of 900 deg. C (INET) and 950 deg. C (AVR). - very high burn-up of 17% FIMA (INET) and 20% FIMA (AVR) which is significantly higher than the license limit of the HTR-Modul (approx. 8% FIMA); it will be explained in this paper why this objective had to be somewhat reduced due to excessive irradiation time requirements and technological difficulties; This paper provides the irradiation history of the experiment performed so far including data on fission gas release. (authors)

  12. 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)

    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

  13. Pebbles and Branching Programs for Tree Evaluation

    CERN Document Server

    Cook, Stephen; Wehr, Dustin; Braverman, Mark; Santhanam, Rahul

    2010-01-01

    We introduce the Tree Evaluation Problem, show that it is in logDCFL (and hence in P), and study its branching program complexity in the hope of eventually proving a superlogarithmic space lower bound. The input to the problem is a rooted, balanced d-ary tree of height h, whose internal nodes are labeled with d-ary functions on [k] = {1,...,k}, and whose leaves are labeled with elements of [k]. Each node obtains a value in [k] equal to its d-ary function applied to the values of its d children. The output is the value of the root. We show that the standard black pebbling algorithm applied to the binary tree of height h yields a deterministic k-way branching program with Theta(k^h) states solving this problem, and we prove that this upper bound is tight for h=2 and h=3. We introduce a simple semantic restriction called "thrifty" on k-way branching programs solving tree evaluation problems and show that the same state bound of Theta(k^h) is tight (up to a constant factor) for all h >= 2 for deterministic thrift...

  14. An economic analysis of fusion breeders

    International Nuclear Information System (INIS)

    This paper presents a study of the economic performance of Fission/Fusion Hybrid devices. This work takes fusion breeder cost estimates and applies methodology and cost factors used in the fission reactor programs to compare fusion breeders with Liquid Metal Fast Breeder Reactors (LMFBR). The results of the analysis indicate that the Hybrid will be in the same competitive range as proposed LMFBRs and have the potential to provide economically competitive power in a future of rising uranium prices. The sensitivity of the results to variations in key parameters is included

  15. A method for estimating maximum static rainfall retention in pebble mulches used for soil moisture conservation

    Science.gov (United States)

    Peng, Hongtao; Lei, Tingwu; Jiang, Zhiyun; Horton, Robert

    2016-06-01

    Mulching of agricultural fields and gardens with pebbles has long been practiced to conserve soil moisture in some semi-arid regions with low precipitation. Rainfall interception by the pebble mulch itself is an important part of the computation of the water balance for the pebble mulched fields and gardens. The mean equivalent diameter (MED) was used to characterize the pebble size. The maximum static rainfall retention in pebble mulch is based on the water penetrating into the pores of pebbles, the water adhering to the outside surfaces of pebbles and the water held between pebbles of the mulch. Equations describing the water penetrating into the pores of pebbles and the water adhering to the outside surface of pebbles are constructed based on the physical properties of water and the pebble characteristics. The model for the water between pebbles of the mulch is based on the basic equation to calculate the water bridge volume and the basic coordination number model. A method to calculate the maximum static rainfall retention in the pebble mulch is presented. Laboratory rain simulation experiments were performed to test the model with measured data. Paired sample t-tests showed no significant differences between the values calculated with the method and the measured data. The model is ready for testing on field mulches.

  16. A novel reusable nanocomposite for complete removal of dyes, heavy metals and microbial load from water based on nanocellulose and silver nano-embedded pebbles.

    Science.gov (United States)

    Suman; Kardam, Abhishek; Gera, Meeta; Jain, V K

    2015-01-01

    The present work proposed a nanocellulose (NC)-silver nanoparticles (AgNPs) embedded pebbles-based composite material as a novel reusable cost-effective water purification device for complete removal of dyes, heavy metals and microbes. NC was prepared using acid hydrolysis of cellulose. The AgNPs were generated in situ using glucose and embedded within the porous concrete pebbles by the technique of inter-diffusion of ion, providing a very strong binding of nanoparticles within the porous pebbles and thus preventing any nanomaterials leaching. Fabrication of a continual running water purifier was achieved by making different layering of NC and Ag nano-embedded pebbles in a glass column. The water purifier exhibited not only excellent dye and heavy metal adsorption capacity, but also long-term antibacterial activity against pathogenic and non-pathogenic bacterial strains. The adsorption mainly occurred through electrostatic interaction and pore diffusion also contributed to the process. The bed column purifier has shown 99.48% Pb(II) and 98.30% Cr(III) removal efficiency along with 99% decontamination of microbial load at an optimum working pH of 6.0. The high adsorption capacity and reusability, with complete removal of dyes, heavy metals and Escherichia coli from the simulated contaminated water of composite material, will provide new opportunities to develop a cost-effective and eco-friendly water purifier for commercial application. PMID:25243917

  17. Breeder reactor fuel fabrication system development

    International Nuclear Information System (INIS)

    Significant progress has been made in the design and development of remotely operated breeder reactor fuel fabrication and support systems (e.g., analytical chemistry). These activities are focused by the Secure Automated Fabrication (SAF) Program sponsored by the Department of Energy to provide: a reliable supply of fuel pins to support US liquid metal cooled breeder reactors and at the same time demonstrate the fabrication of mixed uranium/plutonium fuel by remotely operated and automated methods

  18. Design and analysis of breeding blanket with helium cooled solid breeder for ITER-TBM

    International Nuclear Information System (INIS)

    Test blanket module (TBM) is one of important components in ITER. Some of related blanket technologies of future fusion, such as tritium self-sufficiency, the exaction of high-grade heat, design criteria and safety requirements and environmental impacts, will be demonstrated in ITER-TBM. In ITER device, the three equatorial ports have allocated for TBM testing. China had proposed to develop independently the ITER-TBM with helium cooled solid breeder in 12th meeting of test blanket workgroup (TBWG-12). In this work, the preliminary design and analysis for Chinese HCSB TBM will be carried out. The TBM must be contains the function of the first wall, breeding blanket, shield and structure. Finally, in the period of preliminary investigation, HCSB TBM design adopt modularization concept which is helium as coolant and tritium purge gas, ferritic/martensitic steel as structural material, Lithium orthosilicate (Li4SiO4) as tritium breeder, beryllium pebble as neutron multiplier. TBM is allocated in standard vertical frame port. HCSB TBM consist of first wall, backplate, breeding sub-modules, caps, grid and support plate, and breeding sub-modules is arranged by layout of 2 x 6 in blanket box. In this paper, main components of HCSB TBM will be described in detail, also performance analysis of main components have been completed. (authors)

  19. A European proposal for an ITER water-cooled solid breeder blanket

    International Nuclear Information System (INIS)

    The water-cooled solid breeder blanket concept proposed here aims to replace the shielding blanket for the enhanced performance phase of the international thermonuclear experimental reactor (ITER). The nominal performances are as follows: an average neutron wall load of 1 MW m-2 which corresponds to a fusion power of about 1.5 GW, and an average neutron fluence of 1 MWy m-2. The proposed blanket concept has been designed to accept a power increase of about 30% and power transients up to 3-5 GW for a short time. This blanket concept is based on a breeder inside tube (BIT)-type blanket with poloidal breeding elements made of 316 L-type stainless steel and filled with lithium metazirconate and beryllium pebbles. Inlet and outlet water temperatures of 160 and 200 C have been considered with a medium-pressure cooling system during plasma burn. The diameters of the breeding elements are compatible with the space available in test fission reactor core channels, making in-pile testing, required for blanket development and qualification, easier. A conservative approach using qualified materials, a blanket concept easily testable in fission reactors and on-going mock-up testing, which can be qualified using the blanket test module during the basic performance phase of ITER, will allow the blanket reliability required for the enhanced performance phase to be achieved. (orig.)

  20. Conceptual study of ferromagnetic pebbles for heat exhaust in fusion reactors with short power decay length

    OpenAIRE

    Gierse, N.; Coenen, J W; C. Thomser; Panin, A.; Ch. Linsmeier; Unterberg, B.; Philipps, V

    2015-01-01

    Ferromagnetic pebbles are investigated as high heat flux (q∥) plasma facing components in fusion devices with short power decay length (λq) on a conceptual level. The ability of a pebble concept to cope with high heat fluxes is retained and extended by the acceleration of ferromagnetic pebbles in magnetic fields. An alloying concept suited for fusion application is outlined and the compatibility of ferromagnetic pebbles with plasma operation is discussed. Steel grade 1.4510 is chosen as a ...

  1. 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.

  2. Fabrication of beryllide pebble as advanced neutron multiplier

    International Nuclear Information System (INIS)

    Highlights: • A new beryllide granulation process that combined process with a plasma sintering method for electrode fabrication and a rotating electrode method (REM) for granulation was suggested. • The beryllide electrode fabrication process was investigated for mass production. • As optimized beryllide electrode indicated higher ductility and was sintered at a lower temperature for a shorter time. • It appears to be more able to not only withstand the thermal shock from arc-discharge during granulation but also produce beryllide pebbles on a large scale. • These optimization results can reduce the time for electrode fabrication by 40%, they suggest the possibility of great reductions in time and cost for mass production of beryllide pebbles. - Abstract: Fusion reactors require advanced neutron multipliers with great stability at high temperatures. Beryllium intermetallic compounds, called beryllides such as Be12Ti, are the most promising materials for use as advanced neutron multipliers. However, few studies have been conducted on the development of mass production methods for beryllide pebbles. A granulation process for beryllide needs to have both low cost and high efficiency. To fabricate beryllide pebbles, a new granulation process is established in this research by combining a plasma sintering method for beryllide synthesis and a rotating electrode method using a plasma-sintered electrode for granulation. The fabrication process of the beryllide electrode is investigated and optimized for mass production. The optimized beryllide electrode exhibits higher ductility and can be sintered at a lower temperature for a shorter time, indicating that it is more suitable not only for withstanding the thermal shock from arc-discharge during granulation but also for producing the beryllide pebbles on a large scale. Accordingly, because these optimization results can reduce the time required for electrode fabrication by 40%, they suggest the possibility of

  3. Tracer-pebble movement along a concave river profile: Virtual velocity in relation to grain size and shear stress

    Science.gov (United States)

    Ferguson, R. I.; Wathen, S. J.

    1998-08-01

    Over 1400 tracer pebbles 16-256 mm in diameter were tracked for 2 years in six reaches of Allt Dubhaig, Scotland, a small gravel-bed river along which shear stress and bed surface grain size decrease toward a local base level. Pebble movement was size-selective both within and between reaches. Within reaches the decrease in mean travel distance with increasing grain size is strongest in the coarse tail of the size distribution. Particle shape has a minor secondary effect. A nondimensional grain velocity, averaged over the duration of competent flow, is used to compare different size classes and reaches. Over 90% of its variance is explained by relative grain size and reach Shields stress. The pattern of size selectivity is consistent with single-event tracer results elsewhere, bedload trap data from our distal reach, and the concept of partial mobility. It provides a mechanism for strong downstream fining by selective transport and deposition along rivers in which stress declines toward base level. The nondimensional prediction equation for grain velocity may be of use in other rivers but requires testing.

  4. Fabrication of Li2TiO3 pebbles by a freeze drying process

    International Nuclear Information System (INIS)

    Li2TiO3 pebbles were successfully fabricated by using a freeze drying process. The Li2TiO3 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 Li2TiO3 pebbles is introduced. The processing parameters, such as solid content in the slurry and the conditions of freeze drying and sintering, are also examined

  5. European DEMO BOT solid breeder blanket

    International Nuclear Information System (INIS)

    The BOT (Breeder Outside Tube) Solid Breeder Blanket for a fusion DEMO reactor is presented. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. In the paper the reference blanket design and external loops are described as well as the results of the theoretical and experimental work in the fields of neutronics, thermohydraulics, mechanical stresses, tritium control and extraction, development and irradiation of the ceramic breeder material, beryllium development, ferromagnetic forces caused by disruptions, safety and reliability. An outlook is given on the remaining open questions and on the required R and D program. (orig.)

  6. Accelerator breeder with uranium, thorium target

    International Nuclear Information System (INIS)

    An accelerator breeder, that uses a low-enriched fuel as the target material, can produce substantial amounts of fissile material and electric power. A study of H2O- and D2O-cooled, UO2, U, (depleted U), or thorium indicates that U-metal fuel produces a good fissile production rate and electrical power of about 60% higher than UO2 fuel. Thorium fuel has the same order of magnitude as UO2 fuel for fissile-fuel production, but the generating electric power is substantially lower than in a UO2 reactor. Enriched UO2 fuel increases the generating electric power but not the fissile-material production rate. The Na-cooled breeder target has many advantages over the H2O-cooled breeder target

  7. Improved fuel element for fast breeder reactor

    International Nuclear Information System (INIS)

    The invention, in which the United States Department of Energy has participated as co-inventor, relates to breeder reactor fuel elements, and specifically to such elements incorporating 'getters', hereafter designated as fission product traps. The main object of the invention is the construction of a fast breeder reactor fuel pin, free from local stresses induced in the cladding by reactions with cesium. According to the invention, the fast breeder fuel element includes a cladding tube, sealed at both ends by a plug, and containing a fissile stack and a fertile stack, characterized by the interposition of a cesium trap between the fissile and fertile stacks. The trap is effective at reactor operating temperatures in retaining and separating the cesium generated in the fissile material and preventing cesium reaction with the fertile stack. Depending on the construction method adopted, the trap may consists of a low density titanium oxide or niobium oxide pellet

  8. The role and problems of the breeder

    International Nuclear Information System (INIS)

    World uranium resources are discussed and it is concluded that the period of availability of uranium for use in the present type of nuclear power stations is not much greater than that of oil. The neutron economies of fast and thermal reactors are compared, and the advantages of the breeder for the world uranium economy are demonstrated. The main impediments to the use of the fast breeder are considerations of safety, public acceptance and economics. Fast reactor safety is discussed and the health hazards and possible mis-use of plutonium for terrorism and weapons proliferation are considered. It is widely accepted that the U.K. cannot economically justify the development of the breeder alone and is likely to choose to co-operate with Western Europe. A public enquiry in the U.K. seems certain and would be welcomed by the nuclear industry. (author)

  9. Fast breeder reactors an engineering introduction

    CERN Document Server

    Judd, A M

    1981-01-01

    Fast Breeder Reactors: An Engineering Introduction is an introductory text to fast breeder reactors and covers topics ranging from reactor physics and design to engineering and safety considerations. Reactor fuels, coolant circuits, steam plants, and control systems are also discussed. This book is comprised of five chapters and opens with a brief summary of the history of fast reactors, with emphasis on international and the prospect of making accessible enormous reserves of energy. The next chapter deals with the physics of fast reactors and considers calculation methods, flux distribution,

  10. Coatings for fast breeder reactor components

    International Nuclear Information System (INIS)

    Several types of metallurgical coatings are used in the unique environments of the fast breeder reactor. Most of the coatings have been developed for tribological applications, but some also serve as corrosion barriers, diffusion barriers, or radionuclide traps. The materials that have consistently given the best performance as tribological coatings in the breeder reactor environments have been coatings based on chromium carbide, nickel aluminide, or Tribaloy 700 (a nickel-base hard-facing alloy). Other coatings that have been qualified for limited applications include chromium plating for low temperature galling protection and nickel plating for radionuclide trapping

  11. Comparison analysis of fusion breeder blanket concepts

    International Nuclear Information System (INIS)

    Based on the wide survey, the development status and key issues of fusion breeder blanket concepts are summarized. Two types of blanket concepts, i.e. solid and liquid breeder blanket, were compared and assessed in terms of engineering feasibility, tritium recovery and control, economic and safety aspects, etc. The advantages and disadvantages of the two types of blanket concepts were clarified from the viewpoint of technology realization and development potential. This study may act as a valuable reference for fusion blanket concept selection and design. (authors)

  12. Status and prospects of advanced fissile fuel breeders

    International Nuclear Information System (INIS)

    Fusion--fission hybrid systems, fast breeder systems, and accelerator breeder systems were compared on a common basis using a simple economic model. Electricity prices based on system capital costs only were computed, and were plotted as functions of five key breeder system parameters. Nominally, hybrid system electricity costs were about twenty-five percent lower than fast breeder system electricity costs, and fast breeder system electricity costs were about forty percent lower than accelerator breeder system electricity costs. In addition, hybrid system electricity costs were very insensitive to key parameter variations on the average, fast breeder system electricity costs were moderately sensitive to key parameter variations on the average, and accelerator breeder system electricity costs were the most sensitive to key parameter variations on the average

  13. The development of breeder reactors in the US

    International Nuclear Information System (INIS)

    This article discusses the early history of breeder development in the US, the early history of the fast reactor in the US, changes during the Carter administration, and the development of LMFBR technology. Topics considered include the intermediate-energy plutonium breeder, the molten plutonium breeder, the aqueous homogeneous reactor, the molten-salt reactor, the liquid metal-fueled reactor, electronuclear breeding, the Experimental Breeder Reactor-I, the Experimental Breeder Reactor-II, the Enrico Fermi Reactor, a programmatic change to ceramic fuel, the South East Fast Oxide Reactor, the sodium void coefficient, the 1000-MWe studies of 1964, the 1000-MWe studies of 1967-1969, the FARET design, the Fast Flux Test Facility, the Clinch River Breeder Reactor (CRBR), the gas-cooled fast breeder, the light-water breeder, materials for cladding and duct walls, and reactor safety. It is pointed out that the Congress opposes the construction of the CRBR, while the Reagan administration strongly supports it

  14. A review of fusion breeder blanket technology, part 1

    International Nuclear Information System (INIS)

    This report presents the results of a study of fusion breeder blanket technology. It reviews the role of the breeder blanket, the current understanding of the scientific and engineering bases of liquid metal and solid breeder blankets and the programs now underway internationally to resolve the uncertainities in current knowledge. In view of existing national expertise and experience, a solid breeder R and D program for Canada is recommended

  15. First-order control of surface roughness at three scales: boundary layer dynamics, tracer dispersion and pebble abrasion

    Science.gov (United States)

    Jerolmack, D. J.; Litwin, K. L.; Phillips, C. B.; Martin, R. L.

    2012-12-01

    In many situations it may be appropriate to treat surfaces as smooth and particles as spherical, however here we focus on scenarios in which the roughness of the surface exerts a first-order control on flow and transport dynamics. We describe three vignettes at three different scales: (1) roughness transitions and resulting sediment transport dynamics over ~10-km distance in a desert dune field; (2) reach-scale river bed roughness and its influence on dispersion of tracer particles in bed load; and (3) the control of particle surface roughness on the nature and rate of pebble abrasion. For (1), we show how the abrupt transition from a flat surface to a dune field may be treated as a step increase in the aerodynamic roughness parameter - so long as the spatial scale considered is significantly larger than that of an individual dune. This increase causes a spatial decline in the boundary stress downwind that may be understood using simple boundary layer theory, resulting in a factor of three decrease in the sand flux over a distance of kilometers. For (2), laboratory and field studies of tracer particles in bed load indicate that they undergo short flights separated by long rest periods having a power-law tail - even in steady flows. We hypothesize that for near-threshold transport - which predominates is coarse-grained rivers - particles become trapped in 'wells' produced by surface roughness, and their rest time is controlled by the time for the surface to scour down and release them. Laboratory observations support this hypothesis, while comparison to non-geophysical 'flows' indicates that these dynamics are generic to transport in disordered systems. Finally, for (3) we report laboratory experiments by our group and others showing how abrasion rate decreases with decreasing particle roughness. Geometric models quantitatively support the intuition that locations of high positive curvature on pebble surfaces are more susceptible to abrasion; as they are

  16. Status of the EXOTIC-8 programme and first in-pile results for Li{sub 2}TiO{sub 3} pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Van der Laan, J.G.; Stijkel, M.P. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Conrad, R.

    1998-03-01

    After renewal of the Tritium Measuring Station the HFR is again fully operational for in-pile breeder irradiations. The EXOTIC-8 series has started with first three experiments on June 12, 1997. First in-pile results have been obtained for Li{sub 2}TiO{sub 3}-pebbles supplied by CEA: preliminary analyses indicate satisfactory in-pile behaviour with fast recovery from transient conditions. Five further experiments have been defined which implies that in the present planning EXOTIC-8 is filled completely up to Fall`98 and 2 of 4 positions are occupied up to Spring`99. P.I.E. results will be obtained from Spring`98 onwards. (J.P.N.)

  17. Tritium recovery from ceramic breeder blanket

    International Nuclear Information System (INIS)

    It is known that chemical forms of tritium released from ceramic breeders are T2O and T2. Among issues relevant to the tritium chemical form, tritium inventory is one of the major criteria in the selection of breeder material. The primary purpose of this report is to study the dependence of tritium inventory in a blanket with ceramic solid breeder on the tritium chemical form. In this light, tritium inventory in a Li2O blanket has been evaluated as a function of tritium chemical form under the conditions of the Japanese Fusion Experimental Reactor (FER). It was shown that in a blanket with Li2O as a breeder, which has a strong affinity to water vapor, the inventory due to T2O adsorption becomes quite large. In order to reduce the T2O adsorption inventory, conversion of the tritium chemical form through an isotope exchange reaction with hydrogen added to the sweep gas (T2O + 2 H2 → H2O + 2 HT) has been proposed, and its advantages and problems have been examined. Lithium hydroxide formation and mass transfer, which are considered to be inherent in the Li2O-T2O system and to be critical issues for the feasibility of a Li2O blanket, have been also discussed. (author)

  18. Possible types of breeders with thorium cycle

    International Nuclear Information System (INIS)

    Neutronics calculations of simplified homogeneous reactor models show the possibility that metal-fueled LMFBRs and coated particle fueled gas cooled reactors achieve reactor doubling times of around 10 years with the thorium cycle. Three concepts of gas-cooled thorium cycle breeders are discused. (Author)

  19. Possible types of breeders with thorium cycle

    International Nuclear Information System (INIS)

    Neutronics calculations of simplified homogeneous reactor models show the possibility that metal-fueled LMFBRs and coated particle fueled gas cooled reactors achieve doubling times of around 10 years with the thorium cycle. Three concepts of gas-cooled thorium cycle breeders are discussed. (Author)

  20. Uranium assessment for the Precambrian pebble conglomerates in southeastern Wyoming

    International Nuclear Information System (INIS)

    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

  1. Tritium release from EXOTIC-7 orthosilicate pebbles. Effect of burnup and contact with beryllium during irradiation

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

    EXOTIC-7 was the first in-pile test with {sup 6}Li-enriched (50%) lithium orthosilicate (Li{sub 4}SiO{sub 4}) pebbles and with DEMO representative Li-burnup. Post irradiation examinations of the Li{sub 4}SiO{sub 4} have been performed at the Forschungszentrum Karlsruhe (FZK), mainly to investigate the tritium release kinetics as well as the effect of Li-burnup and/or contact with beryllium during irradiation. The release rate of Li{sub 4}SiO{sub 4} from pure Li{sub 4}SiO{sub 4} bed of capsule 28.1-1 is characterized by a broad main peak at about 400degC and by a smaller peak at about 800degC, and that from the mixed beds of capsule 28.2 and 26.2-1 shows again these two peaks, but most of the tritium is now released from the 800degC peak. This shift of release from low to high temperature may be due to the higher Li-burnup and/or due to contact with Be during irradiation. Due to the very difficult interpretation of the in-situ tritium release data, residence times have been estimated on the basis of the out-of-pile tests. The residence time for Li{sub 4}SiO{sub 4} from caps. 28.1-1 irradiated at 10% Li-burnup agrees quite well with that of the same material irradiated at Li-burnup lower than 3% in the EXOTIC-6 experiment. In spite of the observed shift in the release peaks from low to high temperature, also the residence time for Li{sub 4}SiO{sub 4} from caps. 26.2-1 irradiated at 13% Li-burnup agrees quite well with the data from EXOTIC-6 experiment. On the other hand, the residence time for Li{sub 4}SiO{sub 4} from caps. 28.2 (Li-burnup 18%) is about a factor 1.7-3.8 higher than that for caps. 26.2-1. Based on these data on can conclude that up to 13% Li-burnup neither the contact with beryllium nor the Li-burnup have a detrimental effect on the tritium release of Li{sub 4}SiO{sub 4} pebbles, but at 18% Li-burnup the residence time is increased by about a factor three. (J.P.N.)

  2. Pebbling and Branching Programs Solving the Tree Evaluation Problem

    CERN Document Server

    Wehr, Dustin

    2010-01-01

    We study restricted computation models related to the Tree Evaluation Problem}. The TEP was introduced in earlier work as a simple candidate for the (*very*) long term goal of separating L and LogDCFL. The input to the problem is a rooted, balanced binary tree of height h, whose internal nodes are labeled with binary functions on [k] = {1,...,k} (each given simply as a list of k^2 elements of [k]), and whose leaves are labeled with elements of [k]. Each node obtains a value in [k] equal to its binary function applied to the values of its children, and the output is the value of the root. The first restricted computation model, called Fractional Pebbling, is a generalization of the black/white pebbling game on graphs, and arises in a natural way from the search for good upper bounds on the size of nondeterministic branching programs (BPs) solving the TEP - for any fixed h, if the binary tree of height h has fractional pebbling cost at most p, then there are nondeterministic BPs of size O(k^p) solving the heigh...

  3. Alternate fuel cycles for fast breeder reactors

    International Nuclear Information System (INIS)

    In this contribution to the syllabus for Subgroup 5D, a full range of alternate breeder fuel cycle options is developed and explored as to energy supply capability, resource utilizations, performance characteristics and technical features that pertain to proliferation resistance. Breeding performance information is presented for designs based on Pu/U, Pu/Th, 233 U/U, etc. with oxide, carbide or metal fuel; with lesser emphasis, heterogeneous and homogeneous concepts are presented. A potential proliferation resistance advantage of a symbiotic system of a Pu/U core, Th blanket breeder producing 233 U for utilization in dispersed LWR's is identified. LWR support ratios for various reactor and fuel types and the increase in uranium consumption with higher support ratios are identified

  4. What can fast breeders do for Ontario

    International Nuclear Information System (INIS)

    Fast reactors have the potential of significantly reducing Ontario's demand for natural resources while meeting virtually any requirements for nuclear power this province may have. The breeding efficiency of the fast reactors does not affect the overall uranium consumption of the system to any significant extent. It is, however, an important economic factor in a breeder/burner system. To minimize the resource consumption, the fast reactors should be introduced in Ontario at the onset of the next century. The 'breeder-burner' mix of reactors can effectively reduce the fissile inventory of the whole power system (including the inventory in irradiated fuel storage bays). For the nuclear capacity growth scenarios thought to be applicable in Ontario, the fast reactor systems have about the same or lower requirements for natural uranium as the best (self-sustaining thorium) CANDU cycles. Compared to all other advanced CANDU cycles, the fast reactors yield a substantial resource saving. (auth)

  5. BEATRIX: The international breeder materials exchange

    International Nuclear Information System (INIS)

    The BEATRIX experiment is an IEA-sponsored effort that involves the exchange of solid breeder materials and shared irradiation testing among research groups in several countries. The materials will be tested in both closed capsules (to evaluate material lifetime) and opened capsules (to evaluate purge-flow tritium recovery). Pre- and post-irradiation measurement of thermophysical and mechanical properties will also be carried out

  6. Technological questions of the breeder fuel cycle

    International Nuclear Information System (INIS)

    Since the contributions by the Karlsruhe Nuclear Research Center to the construction of SNR 300 have been completed to a large extent and irradiated KNK II fuel subassemblies have now become available, the possibility and necessity arise of concentrating efforts on the breeder fuel cycle. This work was started in 1980. The 17 papers presented at this seminar will provide a survey of intermediate results obtained until today. (orig./HP)

  7. FOWL CHOLERA IN A BREEDER FLOCK

    OpenAIRE

    Z. Parveen, A. A. Nasir, K.Tasneem and A. Shah

    2003-01-01

    During January, 2003 Pasteurella multocida the causative agent of fowl cholera was isolated from a breeder flock in Lahore District. The age of the flock was 245 days. Increased mortality, swollen wattles and lameness were the clinical findings present in almost all the affected birds, while gross lesions were typical of fowl cholera. To prove the virulence of the organism, mice and six-week old cockerals were infected and P. multocida was reisolated.

  8. The fast breeder reactor Rapsodie (1962)

    International Nuclear Information System (INIS)

    In this report, the authors describe the Rapsodie project, the French fast breeder reactor, as it stands at construction actual start-up. The paper provides informations about: the principal neutronic and thermal characteristics, the reactor and its cooling circuits, the main handling devices of radioactive or contaminated assemblies, the principles and means governing reactor operation, the purposes and locations of miscellaneous buildings. Rapsodie is expected to be critical by 1964. (authors)

  9. Fast Breeder Development: EDF's point of view

    International Nuclear Information System (INIS)

    This paper presents EDF's views and contributions to fast breeder development and to the French SFR trilateral program. Utility requirements are first outlined, based on the approach followed for the EPR reactor. R and D contributions are presented in the areas of core physics, safety, technology innovations, materials, deployment and fuel cycle scenarios. The paper also deals with some of the issues of the 2020 French prototype as seen by EDF.

  10. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  11. The breeder's broke - who will save it

    International Nuclear Information System (INIS)

    In February, Bonn must decide on whether to tear down the most expensive building of the country: The Fast Breeder at Kalkar, which once seemed to lead to a glorious future. DM 3.500 million have been spent on it already and DM 1.000 million more will be needed. But the state has no money. The report is given by Wolfgang Hoffmann and Horst Bieber. (orig.)

  12. Future designs of breeder reactors (Europe, USA)

    International Nuclear Information System (INIS)

    Sodium-cooled reactors with a fast neutron core today are the only fission reactors that offer the reactor physics required for the breeding process and the complete conversion of U-238 or Th-232 into fissile fuel. There are currently five prototype breeder reactors in operation in England, France, and the USSR. The trends observable in development work aim at reducing capital cost, enhancing and improving passive shutdown performance, and simplifying the fuel cycle. (orig.)

  13. BREEDER: a microcomputer program for financial analysis of a large-scale prototype breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giese, R.F.

    1984-04-01

    This report describes a microcomputer-based, single-project financial analysis program: BREEDER. BREEDER is a user-friendly model designed to facilitate frequent and rapid analyses of the financial implications associated with alternative design and financing strategies for electric generating plants and large-scale prototype breeder (LSPB) reactors in particular. The model has proved to be a useful tool in establishing cost goals for LSPB reactors. The program is available on floppy disks for use on an IBM personal computer (or IBM look-a-like) running under PC-DOS or a Kaypro II transportable computer running under CP/M (and many other CP/M machines). The report documents version 1.5 of BREEDER and contains a user's guide. The report also includes a general overview of BREEDER, a summary of hardware requirements, a definition of all required program inputs, a description of all algorithms used in performing the construction-period and operation-period analyses, and a summary of all available reports. The appendixes contain a complete source-code listing, a cross-reference table, a sample interactive session, several sample runs, and additional documentation of the net-equity program option.

  14. BREEDER: a microcomputer program for financial analysis of a large-scale prototype breeder reactor

    International Nuclear Information System (INIS)

    This report describes a microcomputer-based, single-project financial analysis program: BREEDER. BREEDER is a user-friendly model designed to facilitate frequent and rapid analyses of the financial implications associated with alternative design and financing strategies for electric generating plants and large-scale prototype breeder (LSPB) reactors in particular. The model has proved to be a useful tool in establishing cost goals for LSPB reactors. The program is available on floppy disks for use on an IBM personal computer (or IBM look-a-like) running under PC-DOS or a Kaypro II transportable computer running under CP/M (and many other CP/M machines). The report documents version 1.5 of BREEDER and contains a user's guide. The report also includes a general overview of BREEDER, a summary of hardware requirements, a definition of all required program inputs, a description of all algorithms used in performing the construction-period and operation-period analyses, and a summary of all available reports. The appendixes contain a complete source-code listing, a cross-reference table, a sample interactive session, several sample runs, and additional documentation of the net-equity program option

  15. History and evolution of the breeder reactor

    International Nuclear Information System (INIS)

    The concept of the breeder reactor is almost as old as the idea of the nuclear reactor itself. From the very first years following the discovery of nuclear fission, scientists and technicians tried to turn mankind's eternal dream into reality; that is, enjoy an abundant source of energy without using up our raw material reserves. Nuclear energy offered several solutions to realize this dream. One of them, fusion, seemed out of our grasp in the near future. But fission of 235U was possible, and the Manhattan Project soon furnished ample proof of this theory. However, everyone working in this field was conscious of the fact that thermal neutron reactors make very inefficient use of the energy potential contained in natural uranium. The solution was to use in a core sufficiently rich in fissile matter, the excess neutrons to convert the 238U, so poorly used by other types of reactors, into fissile 239Pu. Regeneration, or 'breeding' of fuel, can multiply the energy drawn from a ton of uranium by a factor of 50 to 100. This would enable us to ward off the specter of an energy shortage and the rapid depletion of uranium mines. As early as 1945 in Los Alamos, Enrico Fermi stated: 'The country which first develops a breeder reactor will have a great competitive edge in atomic energy.' The development of the breeder reactor in the USA and around the world is discussed

  16. High temperature chemical compatibility between SiC composites and Be pebbles

    International Nuclear Information System (INIS)

    SiC composite reinforced fibres are still considered to be an important material to be used in nuclear fusion reactors due to their high temperature and low neutron activation properties. Two different kinds of SiC/SiCf composite were manufactured, one of them presenting an extra SiC coating obtained by chemical vapour deposition technique. Several samples of both materials were placed inside a Be pebble bed and the whole set-up annealed at 800 deg. C for 550 h in a reducing atmosphere, simulating fusion reactor conditions. Surface chemical reactions were investigated with nuclear microprobe analyses techniques and complemented with SEM analysis. For the uncoated samples, surface oxidation is accompanied by a strong C depletion and a Be diffusion. Two different behaviours were found for the coated samples. One of those samples showed extended regions where surface was left almost unaltered. The general behaviour, however, was an increase in the number and extension of the cracks already observed at the surface of the coated virgin samples

  17. Transient analysis for the high temperature pebble bed reactor coupled to the energy conversion system

    International Nuclear Information System (INIS)

    This paper describes the results of the calculational coupling between a high temperature reactor code and a thermal hydraulic code for the energy conversion system. This coupling has been developed in order to come to a more detailed and realistic simulation of the entire HTR system. Combining the two codes reduces the number of assumptions that have to be made related to the boundary conditions of the two separate codes. The paper describes the models used for the dynamic components of the energy conversion system, and shows the results of the calculation for two operational transients in order to demonstrate the effects of the interaction between reactor core and its energy conversion system. (author)

  18. Transient analysis of the pebble-bed HTGR with the DSNP simulation language. V.1

    International Nuclear Information System (INIS)

    This is the final report of the first phase of the research contract between KFA-Julich and Soreq NRC covering the period of January 198l to December 1982. During this period most of the components necessary for the dyanmic simulation of the PNP-500 high temperature gas-cooled reactor were developed, tested and included in the DSNP library. Components available in the DSNP simulation language were modified to be applicable to compressible flow systems. The report describes each of these components and provides instructions for their application to system simulation in the DSNP. The primary, intermediate and water-steam heat removal loops were simulated and tested step by step by developing DSNP simulation programs with progressively increasing complexity. The details of each of these simulations is presented together with the partial results obtained. The partial simulations were combined to obtain a full scale PNP-500 simulation. Several accidents were studied with this simulation program particularly a turbine trip, a reactivity accident and a depressurization accident. The major conclusion from the study is that the PNP-500 is inherently a very stable and safe system under a large variety of extreme operating and accident conditions. (author)

  19. Some simulation aspects, from molecular systems to stochastic geometries of pebble bed reactors

    International Nuclear Information System (INIS)

    After a brief presentation of his teaching and supervising activities, the author gives an overview of his research activities: investigation of atoms under high intensity magnetic field (investigation of the electronic structure under these fields), studies of theoretical and numerical electrochemistry (simulation coupling molecular dynamics and quantum calculations, comprehensive simulations of molecular dynamics), and studies relating stochastic geometry and neutron science

  20. South Africa's Pebble Bed Modular Reactor, a new design for our nuclear future

    International Nuclear Information System (INIS)

    Power Utilities will in the future need to look at various means of generating power during the 21st century. The demands regarding new generation are challenged by such issues as; costs, time to construct, the add on safety requirements of present day nuclear power plant designs and the emissions generated by fossil fuels as reflected in the Kyoto Protocol. These challenges are also aligned with the deforestation, land decimation and releases of methane gas caused by the so-called 'clean' Hydro power plants in many parts of the world. Presently South Africa is looking at various generation mixes for the future. Although the demand in South Africa is currently lower than the capacity, it is anticipated that new capacity will have to be commissioned by about 2008. Even the moderate growth of 2,5% (as was experienced in our last fiscal year) will result in peak electricity demand exceeding capacity between 2005 and 2010. In addition, Eskom's older power stations reach the end of their design life after 2025. South Africa will, therefore, need to access and use all natural resources to produce the additional 20,000MW of electricity that will be needed by 2025 this will of course include a nuclear option. Throughout the world, it is noted that, along with the environmental issues affecting power generation the real leading issue is cost. South Africa has one of the lowest power costs in the world, based on its abundant low-cost coal. As with other Eskom low cost options such as, coal fired generation situated at the pit-head and imported hydro, the PBMR costs will have to meet these demanding cost targets set by Eskom's existing power plants. However, PBMR is virtually independent of location and the intention is that PBMR costs will be in the order of US 2,0c/kWh. The costs of decommissioning, long-term storage of radioactive waste and insurance are included in these estimates. This cost per unit of electricity produced would, however, be much lower than a coal-fired plant at the South African coast or the world average cost of US 3,4c/kWh. The paper presented will discuss the PBMR technology designed to meet the various demands and challenges placed upon power producers world wide; that is: - the need for Low Cost, Environmentally Friendly and Safe power production

  1. Technical status of the pebble bed modular reactor (PBMR-SA) conceptual design

    International Nuclear Information System (INIS)

    The reactor study is well underway seen from a broad spectrum of disciplines and technology. The objective power output with a high efficiency direct cycle power conversion unit remains promising after compiling the first critical analysis of the core and the power conversion unit. The stability and controllability of the system are demonstrated by the engineering simulator. The main system and components are basically specified for costing purposes. A first plant layout has been completed demonstrating the positions of main components, personnel movement, installation methods for large components, etc. A cryptic report style presentation includes study objectives, indicating guiding documents, giving an overview of design and analyses work done as well as a few sketches and diagram are included in this paper. Most of these sketches and diagrams are small replicas of large drawings and are therefore not readable but can be used as references. (author)

  2. What determines hatchling weight: breeder age or incubated egg weight?

    OpenAIRE

    AB Traldi; Menten JFM; CS Silva; PV Rizzo; PWZ Pereira; J Santarosa

    2011-01-01

    Two experiments were carried out to determine which factor influences weight at hatch of broiler chicks: breeder age or incubated egg weight. In Experiment 1, 2340 eggs produced by 29- and 55-week-old Ross® broiler breeders were incubated. The eggs selected for incubation weighed one standard deviation below and above average egg weight. In Experiment 2, 2160 eggs weighing 62 g produced by breeders of both ages were incubated. In both experiments, 50 additional eggs within the weight interval...

  3. Challenges in Forming the Solar System's Giant Planet Cores via Pebble Accretion

    CERN Document Server

    Kretke, K A

    2014-01-01

    Though ~10 Earth mass rocky/icy cores are commonly held as a prerequisite for the formation of gas giants, theoretical models still struggle to explain how these embryos can form within the lifetimes of gaseous circumstellar disks. In recent years, aerodynamic-aided accretion of "pebbles," objects ranging from centimeters to meters in size, has been suggested as a potential solution to this long-standing problem. While pebble accretion has been demonstrated to be extremely effective in local simulations that look at the detailed behavior of these pebbles in the vicinity of a single planetary embryo, to date there have been no global simulations demonstrating the effectiveness of pebble accretion in a more complicated, multi-planet environment. Therefore, we have incorporated the aerodynamic-aided accretion physics into LIPAD, a Lagrangian code which can follow the collisional / accretional / dynamical evolution of a protoplanetary system, to investigate the how pebble accretion manifests itself in the larger ...

  4. Production of various sizes and some properties of beryllium pebbles by the rotating electrode method

    Energy Technology Data Exchange (ETDEWEB)

    Iwadachi, T.; Sakamoto, N.; Nishida, K. [NGK Insulators Ltd., Nagoya (Japan); Kawamura, H.

    1998-01-01

    The particle size distribution of beryllium pebbles produced by the rotating electrode method was investigated. Particle size depends on some physical properties and process parameters, which can practicaly be controlled by varying electrode angular velocities. The average particle sizes produced were expressed by the hyperbolic function with electrode angular velocity. Particles within the range of 0.3 and 2.0 mm in diameter are readily produced by the rotating electrode method while those of 0.2 mm in diameter are also fabricable. Sphericity and surface roughness were good in each size of pebble. Grain sizes of the pebbles are 17 {mu} m in 0.25 mm diameter pebbles and 260 {mu} m in 1.8 mm diameter pebbles. (author)

  5. DEM simulation of pebble flow in HTR-10 core by phenomenological method

    International Nuclear Information System (INIS)

    The 10 MW High Temperature Gas-cooled Reactor (HTR-10), developed at Tsinghua University, is an important test advanced reactor in the world. The pebble flow is of fundamental significance for the HTR-10. The discrete element method validated by experiments was used to study pebble flow in the HTR-10 core by the phenomenological method. A 1 : 1 scale computational model to the HTR-10 was utilized to simulate the motion of 27000 spheres, including the flows with different frictional coefficients and base angles. It is found that the pebble flow inside the HTR-10 core is uniform. The stagnant region does not exist. The larger the frictional coefficient or the base angle is, the more uniform the pebble flow is. When the frictional coefficient is 0.8, the HTR-10 maintains a normal discharge operation without stagnant pebbles. This work is important to further optimization of HTR design and development. (authors)

  6. Gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

  7. Liquid metal cooled fast breeder nuclear reactor

    International Nuclear Information System (INIS)

    A liquid metal cooled fast breeder nuclear reactor has a core comprising a plurality of fuel assemblies supported on a diagrid and submerged in a pool of liquid metal coolant within a containment vessel, the diagrid being of triple component construction and formed of a short cylindrical plenum mounted on a conical undershell and loosely embraced by a fuel store carrier. The plenum merely distributes coolant through the fuel assemblies, the load of the assemblies being carried by the undershell by means of struts which penetrate the plenum. The reactor core, fuel store carrier and undershell provide secondary containment for the plenum. (UK)

  8. Large scale breeder reactor pump dynamic analyses

    International Nuclear Information System (INIS)

    The lateral natural frequency and vibration response analyses of the Large Scale Breeder Reactor (LSBR) primary pump were performed as part of the total dynamic analysis effort to obtain the fabrication release. The special features of pump modeling are outlined in this paper. The analysis clearly demonstrates the method of increasing the system natural frequency by reducing the generalized mass without significantly changing the generalized stiffness of the structure. Also, a method of computing the maximum relative and absolute steady state responses and associated phase angles at given locations is provided. This type of information is very helpful in generating response versus frequency and phase angle versus frequency plots

  9. Tritium dynamics in fusion reactor solid breeder

    International Nuclear Information System (INIS)

    In the field of the NET research progrm, the chemical and diffusive processes involved in solid ceramic breeder materials have been analysed. A mathematical model describing the phenomena has been developed to obtain a quantitative evaluation for a first design approach. The data obtained by means of the above mentioned model are in good agreement with the data obtained by other research groups working in Europe and in United States. The computer codes BLANKET2, MC2, FWBC, have been developed to simulate the phenomena

  10. Neutronics design for a fusion breeder

    International Nuclear Information System (INIS)

    As a fusion breeder, one of the most important figure is support ratio which reflects the economic and fuel production performance of the system to a great extent. In this paper, the support ratio is calculated by using one dimension transport program ANISN and optimized by adjusting 6Li enrichment and blanket arrangement. The radial distribution of producted U-233 is also taken into account. Measures are taken for better blanket design, and satisfactory results are obtained. Tritium breeding ratio T reaches 1.11 and the support ratio is enhanced from 11 to 14. The engineering, safety and environment performance are improved

  11. ITER solid breeder blanket materials database

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.C. [Argonne National Lab., IL (United States); Dienst, W. [Kernforschungszentrum Karlsruhe GmbH (Germany). Inst. fuer Material- und Festkoerperforschung; Flament, T. [CEA Centre d`Etudes de Fontenay-aux-Roses (France). Commissariat A L`Energie Atomique; Lorenzetto, P. [NET Team, Garching (Germany); Noda, K. [Japan Atomic Energy Research Inst., Takai, Ibaraki, (Japan); Roux, N. [CEA Centre d`Etudes et de Recherches Les Materiaux (France). Commissariat a L`Energie Atomique

    1993-11-01

    The databases for solid breeder ceramics (Li{sub 2},O, Li{sub 4}SiO{sub 4}, Li{sub 2}ZrO{sub 3} and LiAlO{sub 2}) and beryllium multiplier material are critically reviewed and evaluated. Emphasis is placed on physical, thermal, mechanical, chemical stability/compatibility, tritium, and radiation stability properties which are needed to assess the performance of these materials in a fusion reactor environment. Correlations are selected for design analysis and compared to the database. Areas for future research and development in blanket materials technology are highlighted and prioritized.

  12. Fast breeder reactor fuel reprocessing in France

    International Nuclear Information System (INIS)

    Simultaneous with the effort on fast breeder reactors launched several years ago in France, equivalent investigations have been conducted on the fuel cycle, and in particular on reprocessing, which is an indispensable operation for this reactor. The Rapsodie experimental reactor was associated with the La Hague reprocessing plant AT1 (1 kg/day), which has reprocessed about one ton of fuel. The fuel from the Phenix demonstration reactor is reprocessed partly at the La Hague UP2 plant and partly at the Marcoule pilot facility, undergoing transformation to reprocess all the fuel (TOR project, 5 t/y). The fuel from the Creys Malville prototype power plant will be reprocessed in a specific plant, which is in the design stage. The preliminary project, named MAR 600 (50 t/y), will mobilize a growing share of the CEA's R and D resources, as the engineering needs of the UP3 ''light water'' plant begins to decline. Nearly 20 tonnes of heavy metals irradiated in fast breeder reactors have been processed in France, 17 of which came from Phenix. The plutonium recovered during this reprocessing allowed the power plant cycle to be closed. This power plant now contains approximately 140 fuel asemblies made up with recycled plutonium, that is, more than 75% of the fuel assemblies in the Phenix core

  13. Prototype fast breeder reactor main options

    International Nuclear Information System (INIS)

    Fast reactor programme gets importance in the Indian energy market because of continuous growing demand of electricity and resources limited to only coal and FBR. India started its fast reactor programme with the construction of 40 MWt Fast Breeder Test Reactor (FBTR). The reactor attained its first criticality in October 1985. The reactor power will be raised to 40 MWt in near future. As a logical follow-up of FBTR, it was decided to build a prototype fast breeder reactor, PFBR. Considering significant effects of capital cost and construction period on economy, systematic efforts are made to reduce the same. The number of primary and secondary sodium loops and components have been reduced. Sodium coolant, pool type concept, oxide fuel, 20% CW D9, SS 316 LN and modified 9Cr-1Mo steel (T91) materials have been selected for PFBR. Based on the operating experience, the integrity of the high temperature components including fuel and cost optimization aspects, the plant temperatures are recommended. Steam temperature of 763 K at 16.6 MPa and a single TG of 500 MWe gross output have been decided. PFBR will be located at Kalpakkam site on the coast of Bay of Bengal. The plant life is designed for 30 y and 75% load factor. In this paper the justifications for the main options chosen are given in brief. (author). 2 figs, 2 tabs

  14. Industrial solar breeder project using concentrator photovoltaics

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, R; Wohlgemuth, J; Burkholder, J; Levine, A; Storti, G; Wrigley, C; McKegg, A

    1979-08-01

    The purpose of this program is to demonstrate the use of a concentrating photovoltaic system to provide the energy for operating a silicon solar cell production facility, i.e., to demonstrate a solar breeder. Solarex has proposed to conduct the first real test of the solar breeder concept by building and operating a 200 kW(e) (peak) concentrating photovoltaic system based on the prototype and system design developed during Phase I. This system will provide all of the electrical and thermal energy required to operate a solar cell production line. This demonstration would be conducted at the Solarex Rockville facility, with the photovoltaic array located over the company parking lot and on an otherwise unusable flood plain. Phase I of this program included a comprehensive analysis of the application, prototype fabrication and evaluation, system design and specification, and a detailed plan for Phases II and III. A number of prototype tracking concentrator solar collectors were constructed and operated. Extensive system analysis was performed to design the Phase II system as a stand-alone power supply for a solar cell production line. Finally, a detailed system fabrication proposal for Phase II and an operation and evaluation plan for Phase III were completed. These proposals included technical, management, and cost plans for the fabrication and exercise of the proposed system.

  15. Coincidence measurements of FFTF breeder fuel subassemblies

    International Nuclear Information System (INIS)

    A prototype coincidence counter developed to assay fast breeder reactor fuel was used to measure four fast-flux test facility subassemblies at the Hanford Engineering Development Laboratory in Richland, Washington. Plutonium contents in the four subassemblies ranged between 7.4 and 9.7 kg with corresponding 240Pu-effective contents between 0.9 and 1.2 kg. Large count rates were observed from the measurements, and plots of the data showed significant multiplication in the fuel. The measured data were corrected for deadtime and multiplication effects using established formulas. These corrections require accurate knowledge of the plutonium isotopics and 241Am content in the fuel. Multiplication-corrected coincidence count rates agreed with the expected count rates based on spontaneous fission-neutron emission rates. These measurements indicate that breeder fuel subassemblies with 240Pu-effective contents up to 1.2 kg can be nondestructively assayed using the shift-register electronics with the prototype counters. Measurements using the standard Los Alamos National Laboratory shift-register coincidence electronics unit can produce an assay value accurate to +-1% in 1000 s. The uncertainty results from counting statistics and deadtime-correction errors. 3 references, 8 figures, 8 tables

  16. Improved structural materials for fast breeder reactors

    International Nuclear Information System (INIS)

    Electricity plays a crucial role in the economic development of our country. Coal is the primary fuel for generation of electricity in India as in many other countries. In India, generation of power by nuclear reactors is very important because of (i) availability of large thorium resource, (ii) constraints on setting up of fossil fuel based power plants and (iii) the negligibly small green house gas emissions by nuclear energy. The nuclear programme of the country is being implemented in three stages: (i) pressurized heavy water reactors of the CANDU type, (ii) sodium-cooled fast reactors and (iii) thorium-based reactors. Sodium-cooled fast reactor (SFR) technology is envisioned to make use of the large thorium reserves available. India has undertaken and made rapid strides in developing SFR technology and building of fast reactors for energy generation. A Fast Breeder Test Reactor (FBTR) of 40 MWt is operating successfully for over 25 years at Indira Gandhi Centre for Atomic Research. Based on the design, construction and operational experience, a 500 MWe Prototype Fast Breeder Reactor (PFBR) has been designed indigenously and is in an advanced stage of construction. Its design is being further optimised for enhanced economy with respect to cost of electricity production, for use in commercial reactors. Currently, several R and D programmes are under implementation for the development of new materials required for improved economy of commercial fast reactors

  17. The fast breeder reactor. v. 1

    International Nuclear Information System (INIS)

    The Energy Committee's report was prepared after hearing evidence (the minutes of which are published in Volume II) from the Central Electricity Generating Board, the United Kingdom Atomic Energy Authority and the Department of Energy. Memoranda received from other interested bodies or individuals were also considered and members of the Committee visited fast breeder projects in France, West Germany and Japan. As well as the development of the fast reactors, the economics and timescale were reviewed. The particular case of the fast breeder reactor and proposed fuel reprocessing plant at Dounreay was considered. The main conclusion is that major expenditure on fast reactor programmes can only be justified if there is a potential economic case, i.e. if the fuel cycle costs are lower than for PWRs. This would only be the case if uranium costs increased greatly. It is not considered worthwhile to participate in the European Fast Reactor although this should be reviewed in 1993 and 1997. The Committee agree with the Government's decision to cease funding the PFR in 1994 and endorses the need to regenerate the local economy which will be affected by this decision. (UK)

  18. Solid breeder blanket design and tritium breeding

    International Nuclear Information System (INIS)

    Thermonuclear D-T power plants will have to be tritium self-sufficient. In addition to recovering the energy carried by the fusion neutrons (about 80% of the fusion energy), the blanket of the reactor will thus have to breed tritium to replace that burnt in the fusion process. This paper is an attempt to cover in a concise way the questions of tritium breeding, and the influence of this issue on the design of, and the material selection for, power reactor blanket relying on the use of solid breeder materials. Tritium breeding requirements - to breed one tritium per fusion neutron - are shown to be quite demanding. To meet them, the blanket must incorporate, in addition to a tritium breeding lithium compound, a neutron multiplier so as to compensate for neutron losses. Presently prefered lithium compounds are Li2O, LiAlO2, Li2ZrO3, Li4SiO4. The neutron multiplier considered in most design concepts is beryllium. Furthermore, the blanket must be designed with a view to minimizing these neutron losses (search for compactness and high coverage ratio of the plasma while minimizing the amount of structures and coolant). The design guidelines are justified and the technological problems which limit their implementation are discussed and illustrated with typical designs of solid breeder blanket. (orig.)

  19. Fabrication of Li{sub 2}TiO{sub 3} pebbles by a freeze drying process

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang-Jin, E-mail: lee@mokpo.ac.kr [Department of Advanced Materials Science and Engineering, Mokpo National University, Muan 534-729 (Korea, Republic of); Park, Yi-Hyun [National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of); Yu, Min-Woo [Department of Advanced Materials Science and Engineering, Mokpo National University, Muan 534-729 (Korea, Republic of)

    2013-11-15

    Li{sub 2}TiO{sub 3} pebbles were successfully fabricated by using a freeze drying process. The Li{sub 2}TiO{sub 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{sub 2}TiO{sub 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.

  20. Mechanical compression tests of beryllium pebbles after neutron irradiation up to 3000 appm helium production

    Energy Technology Data Exchange (ETDEWEB)

    Chakin, V., E-mail: vladimir.chakin@kit.edu [Karlsruhe Institute of Technology, Institite for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rolli, R.; Moeslang, A. [Karlsruhe Institute of Technology, Institite for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain)

    2015-04-15

    Highlights: • Compression tests of highly neutron irradiated beryllium pebbles have been performed. • Irradiation hardening of beryllium pebbles decreases the steady-state strain-rates. • The steady-state strain-rates of irradiated beryllium pebbles exceed their swelling rates. - Abstract: Results: of mechanical compression tests of irradiated and non-irradiated beryllium pebbles with diameters of 1 and 2 mm are presented. The neutron irradiation was performed in the HFR in Petten, The Netherlands at 686–968 K up to 1890–2950 appm helium production. The irradiation at 686 and 753 K cause irradiation hardening due to the gas bubble formation in beryllium. The irradiation-induced hardening leads to decrease of steady-state strain-rates of irradiated beryllium pebbles compared to non-irradiated ones. In contrary, after irradiation at higher temperatures of 861 and 968 K, the steady-state strain-rates of the pebbles increase because annealing of irradiation defects and softening of the material take place. It was shown that the steady-state strain-rates of irradiated beryllium pebbles always exceed their swelling rates.

  1. Activation Calculation for a Fusion Experimental Breeder FEB-E

    Institute of Scientific and Technical Information of China (English)

    FENGKaiming

    2002-01-01

    A fusion breeder might be an essential intermediate application of fusion energy at earlier term, since it has the potential to provide plenty of commercial fissile fuel. Based on fusion physics and technologies available at present and in the near future, the realistic fusion experimental breeder, FEB-E was designed.

  2. Uranium deposits in Proterozoic quartz-pebble conglomerates

    International Nuclear Information System (INIS)

    This report is the result of an effort to gather together the most important information on uranium deposits in Proterozoic quartz-pebble conglomerates in the United States of America, Canada, Finland, Ghana, South Africa and Australia. The paper discusses the uranium potential (and in some cases also the gold potential in South Africa, Western Australia and Ghana) in terms of ores, sedimentation, mineralization, metamorphism, placers, geologic formations, stratigraphy, petrology, exploration, tectonics and distribution. Geologic history and application of geologic models are also discussed. Glacial outwash and water influx is also mentioned. The uranium deposits in a number of States in the USA are covered. The Witwatersrand placers are discussed in several papers. Refs, figs, tabs

  3. Accelerator breeder nuclear fuel production: concept evaluation of a modified design for ORNL's proposed TME-ENFP

    International Nuclear Information System (INIS)

    Recent advances in accelerator beam technology have made it possible to improve the target/blanket design of the Ternary Metal Fueled Electronuclear Fuel Producer (TMF-ENFP), an accelerator-breeder design concept proposed by Burnss et al. for subcritical breeding of the fissile isotope 233U. In the original TMF-ENFP the 300-mA, 1100-MeV proton beam was limited to a small diameter whose power density was so high that a solid metal target could not be used for producing the spallation neutrons needed to drive the breeding process. Instead the target was a central column of circulating liquid sodium, which was surrounded by an inner multiplying region of ternary fuel rods (239Pu, 232Th, and 238U) and an outer blanket region of 232Th rods, with the entire system cooled by circulating sodium. In the modified design proposed here, the proton beam is sufficiently spread out to allow the ternary fuel to reside directly in the beam and to be preceded by a thin (nonstructural) V-Ti steel firThe spread beam mandated a change in the design configuration (from a cylindrical shape to an Erlenmeyer flask shape), which, in turn, required that the fuel rods (and blanket rods) be replaced by fuel pebbles. The fuel residence time in both systems was assumed to be 90 full power days. A series of parameter optimization calculations for the modified TMF-ENFP led to a semioptimized system in which the initial 239Pu inventory of the ternary fuel was 6% and the fuel pebble diameter was 0.5 cm. With this system the 233Pu production rate of 5.8 kg/day reported for the original TMF-ENFP was increased to 9.3 kg/day, and the thermal power production at beginning of cycle was increased from 3300 MW(t) to 5240 MW(t). 31 refs., 32 figs., 6 tabs

  4. Water chemistry of breeder reactor steam generators

    International Nuclear Information System (INIS)

    The water quality requirements will be described for breeder reactor steam generators, as well as specifications for balance of plant protection. Water chemistry details will be discussed for the following power plant conditions: feedwater and recirculation water at above and below 5% plant power, refueling or standby, makeup water, and wet layup. Experimental data will be presented from tests which included a departure from nucleate boiling experiment, the Few Tube Test, with a seven tube evaporator and three tube superheater, and a verification of control and on-line measurement of sodium ion in the ppB range. Sampling and instrumentation requirements to insure adherence to the specified water quality will be described. Evaporator cleaning criteria and data from laboratory testing of chemical cleaning solutions with emphasis on flow, chemical composition, and temperature will be discussed

  5. Safeguards in Prototype Fast Breeder Reactor Monju

    International Nuclear Information System (INIS)

    The assemblies loaded in the core and stored in the ex-vessel storage tank (EVST) are in liquid sodium in the Japanese prototype fast breeder reactor (FBR) Monju. Since it is difficult to apply a direct verification procedure for the fuel assemblies in these areas, a dual containment and surveillance system consisting of two monitoring devices such as surveillance camera and radiation monitor that are functionally independent has been applied. In addition, the Monju Remote Monitoring System was developed to strengthen the continuous surveillance and to reduce the load of the inspection activities. Furthermore, the ex-vessel transfer machine radiation monitor (EVRM) and the exit gate monitor (EXGM) were upgraded to strengthen the monitoring of spent blanket fuel assemblies and to improve the reliability of distinguishing between fuel assemblies and non-fuel items. As the result, the integrated safeguards was introduced in November 2009, and the effective safeguards activities have been implemented in Monju. (author)

  6. Lithium reprocessing technology for ceramic breeders

    Science.gov (United States)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Saito, Minoru; Tatenuma, Katuyashi; Kainose, Mitsuru

    1995-03-01

    Lithium ceramics have been receiving considerable attention as tritium breeding materials for fusion reactors. Reprocessing technology development for these materials is proposed to recover lithium, as an effective use of resources and to remove radioactive isotopes. Four potential ceramic breeders (Li 2O, LiAlO 2, Li 2ZrO 3 and Li 4SiO 4) were prepared in order to estimate their dissolution properties in water and various acids (HCl, HNO 3, H 2SO 4, HF and aqua regia). The dissolution rates were determined by comparing the weight of the residue with that of the starting powder (the weight method). Recovery properties of lithium were examined by the precipitation method.

  7. Development of Solid Breeder Blanket at JAERI

    International Nuclear Information System (INIS)

    Japan Atomic Energy Research Institute (JAERI) has been performing blanket development based on the long-term research program of fusion blankets in Japan, which was approved by the Fusion Council of Japan in 1999. The blanket development consists of out-pile R and D, In-pile R and D, TBM Neutronics and TPR Tests and Tritium Recovery System R and D. Based on the achievements of element technology development, the R and D program is now stepping to the engineering testing phase, in which scalable mockup tests will be performed for obtaining engineering data unique to the specific structure of the components, with the objective to define the fabrication specification of test blanket modules for ITER. This paper presents the major achievements of the element technology development of solid breeder blanket in JAERI

  8. Who cracked the pebbles in the gravel pit - lithostatic pressure or a bunch of faults?

    Science.gov (United States)

    Tuitz, Christoph; Exner, Ulrike; Grasemann, Bernhard; Preh, Alexander

    2010-05-01

    The occurrence of radially, brittle fractured pebbles from unconsolidated sediments were investigated in a gravel pit south of St. Margarethen (Burgenland, Austria). The outcrop is located in the Neogene Eisenstadt-Sopron Basin, which is a sub-basin on the SE border of the Vienna Basin. The sediments, which were deposited during the Sarmatian and Pannonian (12.7-7.2 Ma), represent a succession of deltaic gravels with intercalations of shallow-marine calcareous sands. Extensional tectonics in these sediments resulted in the generation of conjugate sets of predominately WNW- and subordinate ESE-dipping normal faults (shear deformation bands). These faults were primarily localized in meter-thick gravel layers and, with increasing displacement, eventually cross-cut other lithologies. The gravel layers contain a significant number of cracked pebbles. Detailed structural mapping of the distribution of cracked pebbles revealed their preferential occurrence in the vicinity of the normal faults and, in these, within zones of roughly uniform-sized pebbles. The findings indicated a strong relation to the mechanics of faulting within the sediment. To find the controlling factors for the localization of pebble fracturing, the grain-size distribution and shape and the number of point contacts of the pebbles were statistically measured. Furthermore, on the basis of point load tests, a breakage criterion was statistically defined which characterizes the breakage behaviour of the pebbles. The results were then used as input parameters for numerical modelling. The Discrete Element Method was applied to simulate the effect of overburden on a certain volume of particles (i.e. the pebbles). In numerical uniaxial compression simulations, the magnitude and the distribution of contact forces between the particles were monitored during compressive loading and repetitively compared with the breakage criterion. If a particle in the simulation reached the criterion, it was automatically

  9. Pebble heater suppresses synthesis of dioxins and furans in off-gas generated by incineration of halogen-rich fuel from WEEE

    Energy Technology Data Exchange (ETDEWEB)

    Schlummer, M.; Gruber, L.; Maeurer, A.; Wolz, G. [Fraunhofer Institute for Process Engineering and Packaging IVV, Freising (Germany); Fischer, W.; Quicker, P. [ATZ-EVUS, Development Center for Process Engineering, Sulzbach-Rosenberg (Germany)

    2004-09-15

    Changes in German and European legislation have led to altered approaches for the disposal of polymer-rich shredding residues (SR). Whereas disposal in landfills was the strategy of choice in the last decades, thermal treatment is supported now. However, when waste electric and electronic equipment (WEEE) is the source of SR, thermal treatment is complicated by a bromine and chlorine load in the lower percent range the presence of polybrominated dioxins and furans (PBDD/F) in the ppb range and by brominated flame retardants including polybrominated biphenyl ethers, which serve as dioxin precursors. Here we present data of a pilot application of the pebble heater technology for the treatment of raw gas derived from the incineration of polymeric materials from WEEE. Since the pilot experiments were performed on an existing pebble heater test plant in the small-technical scale, waste throughput and experimental design had to be adjusted to the given circumstances. As the study focussed on exhaust treatment and not on the incineration process itself, a liquid fuel was applied as a model for SR from WEEE. The incineration of a liquid fuel was preferred, since it could be implemented in the given test plant by spray injection, thus minimising technical modifications of the test plant and optimising the combustion efficiency compared to incineration of solid polymer granulates. Fuel and exhaust gases, which passed the pebble heater bed, were sampled and analysed for PCDD/F and PBDD/F. The pilot incineration was tested for the compliance with the PCDD/F emission limits given by European directive 2000/76/EC, and overall mass balances were calculated for PCDD/F and PBDD/F.

  10. A comparison of fusion breeder/fission client and fission breeder/fission client systems for electrical energy production

    International Nuclear Information System (INIS)

    A parametric study that evaluated the economic performance of breeder/client systems is described. The linkage of the breeders to the clients was modelled using the stockpile approach to determine the system doubling time. Since the actual capital costs of the breeders are uncertain, a precise prediction of the cost of a breeder was not attempted. Instead, the breakeven capital cost of a breeder relative to the capital cost of a client reactor was established by equating the cost of electricity from the breeder/client system to the cost of a system consisting of clients alone. Specific results are presented for two breeder/client systems. The first consisted of an LMFBR with LWR clients. The second consisted of a DT fusion reactor (with a 238U fission suppressed blanket) with LWR clients. The economics of each system was studied as a function of the cost of fissile fuel from a conventional source. Generally, the LMFBR/LWR system achieved relatively small breakeven capital cost ratios; the maximum ratio computed was 2.2 (achieved at approximately triple current conventional fissile material cost). The DTFR/LWR system attained a maximum breakeven capital cost ratio of 4.5 (achieved at the highest plasma quality (ignited device) and triple conventional fissile cost)

  11. Light-water breeder reactors: preliminary safety and environmental information document. Volume III

    International Nuclear Information System (INIS)

    Information is presented concerning prebreeder and breeder reactors based on light-water-breeder (LWBR) Type 1 modules; light-water backfit prebreeder supplying advanced breeder; light-water backfit prebreeder/seed-blanket breeder system; and light-water backfit low-gain converter using medium-enrichment uranium, supplying a light-water backfit high-gain converter

  12. Research and development status of ceramic breeder materials

    International Nuclear Information System (INIS)

    The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction, both of which are critical to development of fusion power. The lithium ceramics continue to show promise as candidate breeder materials. This promise was also recognized by the International Thermonuclear Experimental Reactor (ITER) design team in its selection of ceramics as the first option breeder material. Blanket design studies have indicated areas in the properties data base that need further investigation. Current studies are focusing on issues such as tritium release behavior at high burnup, changes in thermophysical properties with burnup, compatibility between ceramic breeder and beryllium multiplier, and phase changes with burnup. Laboratory and in-reactor tests are underway, some as part of an international collaboration for development of ceramic breeder materials. 36 refs

  13. Neutronic optimization of solid breeder blankets for STARFIRE design

    International Nuclear Information System (INIS)

    Extensive neutronic tradeoff studies were carried out to define and optimize the neutronic performance of the different solid breeder options for the STARFIRE blanket design. A set of criteria were employed to select the potential blanket materials. The basic criteria include the neutronic performance, tritium-release characteristics, material compatibility, and chemical stability. Three blanket options were analyzed. The first option is based on separate zones for each basic blanket function where the neutron multiplier is kept in a separate zone. The second option is a heterogeneous blanket type with two tritium breeder zones. In the first zone the tritium breeder is assembled in a neutron multiplier matrix behind the first wall while the second zone has a neutron moderator matrix instead of the neutron multiplier. The third blanket option is similar to the second concept except the tritium breeder and the neutron multiplier form a homogeneous mixture

  14. Fabrication, properties, and tritium recovery from solid breeder materials

    International Nuclear Information System (INIS)

    The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction, both of which are critical to development of fusion power. The lithium ceramics continue to show promise as candidate breeder materials. This promise was recognized by the International Thermonuclear Experimental Reactor (ITER) design team in its selection of ceramics as the first option for the ITER breeder material. Blanket design studies have indicated properties in the candidate materials data base that need further investigation. Current studies are focusing on tritium release behavior at high burnup, changes in thermophysical properties with burnup, compatibility between the ceramic breeder and beryllium multiplier, and phase changes with burnup. Laboratory and in-reactor tests, some as part of an international collaboration for development of ceramic breeder materials, are underway. 133 refs., 1 fig

  15. Simulation of physical parameter for in-pipe tritium breeder

    International Nuclear Information System (INIS)

    It is necessary to build in-pipe tritium breeder in our country in order to assess breeder material of tritium breeder module (TBM) and to find the release law of tritium. The irradiation vessel is one of the key components of TBM. The physical parameters about in-pipe tritium breeder were simulated with MCNP code. The values of the self-shielding factor, equivalent cross-section, daily production of tritium and total heating power are separately 0.435, 1.09 x 10-22 cm2, 2.8 x 1010 Bq and 8.2 kW. And they would provide necessary data for designing the irradiation vessel. (authors)

  16. Fast-breeder-power reactor records in the INIS database

    International Nuclear Information System (INIS)

    This report presents a statistical analysis of more than 19,700 records of publications concerned with research and technology in the field of fast breeder power fission reactors which are included in the INIS Bibliographic Database for the period from 1970. to 1999. The main objectives of this bibliometric study were: to make an inventory of the fast breeder power reactor related records in the INIS Database; to provide statistics and scientific indicators for the INIS users, namely science managers, researchers, engineers, operators, scientific editors and publishers, decision-makers in the field of fast breeder power reactors related subjects; to extract other useful information from the INIS Bibliographic Database about articles published in fast breeder reactors research and technology. The quantitative data in this report are obtained for various properties of relevant INIS records such as year of publication, secondary subject categories, countries of publication, language, publication types, literary types, etc. (author)

  17. Research about the Influence of Environmental Factors on Breeders Quality

    Directory of Open Access Journals (Sweden)

    Adina Popescu

    2011-10-01

    Full Text Available Along the growth period of the breeders, the monitoring of environmental parameters is a fundamental condition toensure the quality of the breeders used for reproduction. The results from the research presented in this paper wereobtained following experimental type investigations developed in vegetation and cold season within Carja 1-Vasluifish farm, on chemical and biological samples which were analyzed within the research laboratory of the Departmentof Aquaculture, Environmental Science and Cadastre. Were analyzed parameters which influence bio-productivity:temperature, oxygen, pH, the concentration of nitrites, nitrates, phosphates, the density and abundance ofphytoplankton and zooplankton, the individual weight and health condition of breeders. Analyzed parametersincluded mean values recorded in the optimal range for fish waters, as reflected in the numerical density andabundance of plankton and the average weight of Asian cyprinids breeders with a plankton nutritional spectrum.

  18. Reprocessing of fast breeder reactor fuels in France

    International Nuclear Information System (INIS)

    The reprocessing of breeder reactor fuels is a direct technical descendant of the reprocessing of thermal reactor fuels which was developped first. The process used is in both cases the PUREX process, which consists in dissolution by nitric acid followed by selective extraction using TBP. In France, the application of this technique to breeder reactor fuels greatly benefited from the scientific and industrial experience initially acquired with metallic fuels of the MAGNOX type and then with oxide fuels of the LWR type

  19. Development of Liquid Type Breeder Technology for ITER-TBM

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Ki Sok; Hong, Bong Geun; Lee, Dong Won

    2008-07-15

    In relation to liquid type TBM technology development, various works are performed. We established a test loop concept to test the MHD effects and materials compatibility for the Pb-17Li breeder material. For the loop construction, electromagnetic pump and storage tank for the Pb-17Li loop was manufactured and some technical requirements are summarised. As a reference, technical literatures relevant to the liquid type TBM materials and the tritium extraction from breeder materials are also surveyed.

  20. Exploding the myths about the fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Burns, S.

    1979-01-01

    This paper discusses the facts and figures about the effects of conservation policies, the benefits of the Clinch River Breeder Reactor demonstration plant, the feasibility of nuclear weapons manufacture from reactor-grade plutonium, diversion of plutonium from nuclear plants, radioactive waste disposal, and the toxicity of plutonium. The paper concludes that the U.S. is not proceeding with a high confidence strategy for breeder development because of a variety of false assumptions.