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Sample records for helium cooled pebble

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

    Wang, C.; Ballinger, R.G.; Stahle, P.W.; Demetri, E.; Koronowski, M.

    2002-01-01

    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)

  2. Helium-cooled pebble bed test blanket module alternative design and fabrication routes

    Lux, M.

    2007-01-01

    According to first results of the recently started European DEMO study, a new blanket integration philosophy was developed applying so-called multi-module segments. These consist of a number of blanket modules flexibly mounted onto a common vertical manifold structure that can be used for replacing all modules in one segment at one time through vertical remote-handling ports. This principle gives new freedom in the design choices applied to the blanket modules itself. Based on the alternative design options considered for DEMO also the ITER test blanket module was newly analyzed. As a result of these activities it was decided to keep the major principles of the reference design like stiffening grid, breeder unit concept and perpendicular arrangement of pebble beds related to the First Wall because of the very positive results of thermo-mechanical and neutronics studies. The present paper gives an overview on possible further design optimization and alternative fabrication routes. One of the most significant improvements in terms of the hydraulic performance of the Helium cooled reactor can be reached with a new First Wall concept. That concept is based on an internal heat transfer enhancement technique and allows drastically reducing the flow velocity in the FW cooling channels. Small ribs perpendicular to the flow direction (transverse-rib roughness) are arranged on the inner surface of the First Wall cooling channels at the plasma side. In the breeder units cooling plates which are mostly parallel but bent into U-shape at the plasma-side are considered. In this design all flow channels are parallel and straight with the flow entering on one side of the parallel plate sections and exiting on the other side. The ceramic pebble beds are embedded between two pairs of such type of cooling plates. Different modifications could possibly be combined, whereby the most relevant discussed in this paper are (i) rib-cooled First Wall channels, (ii) U-bent cooling plates for

  3. Manufacturing aspects in the design of the breeder unit for Helium Cooled Pebble Bed blankets

    Rey, J.; Ihli, T.; Filsinger, D.; Polixa, C.

    2007-01-01

    The breeding blanket programme has been in the focus of European fusion research for more than a decade. Recently, it has been driven by the EU Power Plant Conceptual Study (PPCS), investigating the potential of fusion energy in a future economic environment. On the way to the first commercial nuclear fusion reactor (DEMO) new studies for reactor in-vessel components have been initiated. One central focus is the design and manufacturing of the blankets that have to ensure the breeding process to maintain the fuel cycle and are also responsible for the extraction of the main part of the reactor heat for power generation. Two kinds are established: One is the Helium Cooled Pebble Bed (HCPB) and the other the Helium Cooled Liquid Lead (HCLL) blanket. Both designs employ three different cooling plate assemblies. The outer, cooled U-shaped shell, namely the First Wall (FW), with two caps builds the blanket box. The structural strength of the blanket box is realized by integrating Stiffening Grids (SG) that separate the equally spaced Breeder Unit (BU) and allow the box, in case of faulted conditions, to withstand an internal pressure of 8 MPa. The cooled SG constitute the side walls of the BU and are also cooled. The BU consists of a dedicated Cooling Plate (CP) assembly. In present studies about the fabrication of Cooling Plates two kinds of diffusion welding processes are focused on. One is based on a Hot Isostatic Gas Process (HIP). The second is a uni-axial Diffusion Welding Process (DWP). In both cases the bond between the two halves of the cooling plate structure is reached by controlled pressure and heat cycles. Approaching larger, realistic scaled components the uncertainty of ensuring uniform process parameters across the bonding zone increases the risk of defect sources and, therefore, makes it difficult to guarantee the required bonding penetration. This study presents an alternative manufacturing strategy. The premises for this strategy are the reduction of

  4. European Helium Cooled Pebble Bed (HCPB) test blanket. ITER design description document. Status 1.12.1996

    Albrecht, H.; Boccaccini, L.V.; Dalle Donne, M.; Fischer, U.; Gordeev, S.; Hutter, E.; Kleefeldt, K.; Norajitra, P.; Reimann, G.; Ruatto, P.; Schleisiek, K.; Schnauder, H.

    1997-04-01

    The Helium Cooled Pebble Bed (HCPB) blanket is based on the use of separate small lithium orthosilicate and beryllium pebble beds placed between radial toroidal cooling plates. The cooling is provided by helium at 8 MPa. The tritium produced in the pebble beds is purged by the flow of helium at 0.1 MPa. The structural material is martensitic steel. It is foreseen, after an extended R and D work, to test in ITER a blanket module based on the HCPB design, which is one of the two European proposals for the ITER Test Blanket Programme. To facilitate the handling operation the Blanket Test Module (BTM) is bolted to a surrounding water cooled frame fixed to the ITER shield blanket back plate. For the design of the test module, three-dimensional Monte Carlo neutronic calculations and thermohydraulic and stress analyses for the operation during the Basic Performance Phase (BPP) and during the Extended Performance Phase (EPP) of ITER have been performed. The behaviour of the test module during LOCA and LOFA has been investigated. Conceptual designs of the required ancillary loops have been performed. The present report is the updated version of the Design Description Document (DDD) for the HCPB Test Module. It has been written in accordance with a scheme given by the ITER Joint Central Team (JCT) and accounts for the comments made by the JCT to the previous version of this report. This work has been performed in the framework of the Nuclear Fusion Project of the Forschungszentrum Karlsruhne and it is supported by the European Union within the European Fusion Technology Program. (orig.) [de

  5. Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

    Panayotov, Dobromir; Grief, Andrew; Merrill, Brad J.; Humrickhouse, Paul; Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon; Poitevin, Yves; Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard

    2016-01-01

    Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

  6. Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

    Panayotov, Dobromir, E-mail: dobromir.panayotov@f4e.europa.eu [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Grief, Andrew [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Merrill, Brad J.; Humrickhouse, Paul [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID (United States); Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Poitevin, Yves [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom)

    2016-11-01

    Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

  7. Thermal-hydraulic calculation and analysis on helium cooled ceramic breeder pebble bed assembly for in-pile irradiation and in-situ tritium extraction

    Guo Chunqiu; Xie Jiachun; Liu Xingmin

    2013-01-01

    In-pile irradiation and in-situ tritium extraction experiment is one of associated domestic research projects in ITER special program. According to the technical requirements of in-pile irradiation experiment of helium cooled ceramic breeder (ceramic) pebble bed assembly in a research reactor, the feasibility of the design for the in-pile irradiation and in-situ tritium extraction experiment of ceramic pebble bed assembly was evaluated. By conducting thermal-hydraulic design calculation with different in-pile irradiation channels, locations and structure parameters for ceramic pebble bed assembly, a reasonable design scheme of ceramic pebble bed assembly satisfying the design requirements for in-pile irradiation was obtained. (authors)

  8. Sensisivity and Uncertainty analysis for the Tritium Breeding Ratio of a DEMO Fusion reactor with a Helium cooled pebble bed blanket

    Nunnenmann, Elena; Fischer, Ulrich; Stieglitz, Robert

    2016-01-01

    An uncertainty analysis was performed for the tritium breeding ratio (TBR) of a fusion power plant of the European DEMO type using the MCSEN patch to the MCNP Monte Carlo code. The breeding blanket was of the type Helium Cooled Pebble Bed (HCPB), currently under development in the European Power Plant Physics and Technology (PPPT) programme for a fusion power demonstration reactor (DEMO). A suitable 3D model of the DEMO reactor with HCPB blanket modules, as routinely used for blanket design c...

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

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

    2014-10-15

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

  10. Experimental investigations of flow distribution in coolant system of Helium-Cooled-Pebble-Bed Test Blanket Module

    Ilić, M.; Schlindwein, G., E-mail: georg.schlindwein@kit.edu; Meyder, R.; Kuhn, T.; Albrecht, O.; Zinn, K.

    2016-02-15

    Highlights: • Experimental investigations of flow distribution in HCPB TBM are presented. • Flow rates in channels close to the first wall are lower than nominal ones. • Flow distribution in central chambers of manifold 2 is close to the nominal one. • Flow distribution in the whole manifold 3 agrees well with the nominal one. - Abstract: This paper deals with investigations of flow distribution in the coolant system of the Helium-Cooled-Pebble-Bed Test Blanket Module (HCPB TBM) for ITER. The investigations have been performed by manufacturing and testing of an experimental facility named GRICAMAN. The facility involves the upper poloidal half of HCPB TBM bounded at outlets of the first wall channels, at outlet of by-pass pipe and at outlets of cooling channels in breeding units. In this way, the focus is placed on the flow distribution in two mid manifolds of the 4-manifold system: (i) manifold 2 to which outlets of the first wall channels and inlet of by-pass pipe are attached and (ii) manifold 3 which supplies channels in breeding units with helium coolant. These two manifolds are connected with cooling channels in vertical/horizontal grids and caps. The experimental facility has been built keeping the internal structure of manifold 2 and manifold 3 exactly as designed in HCPB TBM. The cooling channels in stiffening grids, caps and breeding units are substituted by so-called equivalent channels which provide the same hydraulic resistance and inlet/outlet conditions, but have significantly simpler geometry than the real channels. Using the conditions of flow similarity, the air pressurized at 0.3 MPa and at ambient temperature has been used as working fluid instead of HCPB TBM helium coolant at 8 MPa and an average temperature of 370 °C. The flow distribution has been determined by flow rate measurements at each of 28 equivalent channels, while the pressure distribution has been obtained measuring differential pressure at more than 250 positions. The

  11. Sensitivity and uncertainty analysis for the tritium breeding ratio of a DEMO fusion reactor with a helium cooled pebble bed blanket

    Nunnenmann Elena

    2017-01-01

    Full Text Available An uncertainty analysis was performed for the tritium breeding ratio (TBR of a fusion power plant of the European DEMO type using the MCSEN patch to the MCNP Monte Carlo code. The breeding blanket was of the type Helium Cooled Pebble Bed (HCPB, currently under development in the European Power Plant Physics and Technology (PPPT programme for a fusion power demonstration reactor (DEMO. A suitable 3D model of the DEMO reactor with HCPB blanket modules, as routinely used for blanket design calculations, was employed. The nuclear cross-section data were taken from the JEFF-3.2 data library. For the uncertainty analysis, the isotopes H-1, Li-6, Li-7, Be-9, O-16, Si-28, Si-29, Si-30, Cr-52, Fe-54, Fe-56, Ni-58, W-182, W-183, W-184 and W-186 were considered. The covariance data were taken from JEFF-3.2 where available. Otherwise a combination of FENDL-2.1 for Li-7, EFF-3 for Be-9 and JENDL-3.2 for O-16 were compared with data from TENDL-2014. Another comparison was performed with covariance data from JEFF-3.3T1. The analyses show an overall uncertainty of ± 3.2% for the TBR when using JEFF-3.2 covariance data with the mentioned additions. When using TENDL-2014 covariance data as replacement, the uncertainty increases to ± 8.6%. For JEFF-3.3T1 the uncertainty result is ± 5.6%. The uncertainty is dominated by O-16, Li-6 and Li-7 cross-sections.

  12. Cooling with Superfluid Helium

    Lebrun, P; Tavian, L [European Organization for Nuclear Research, Geneva (Switzerland)

    2014-07-01

    The technical properties of helium II (‘superfluid’ helium) are presented in view of its applications to the cooling of superconducting devices, particularly in particle accelerators. Cooling schemes are discussed in terms of heat transfer performance and limitations. Large-capacity refrigeration techniques below 2 K are reviewed, with regard to thermodynamic cycles as well as process machinery. Examples drawn from existing or planned projects illustrate the presentation. Keywords: superfluid helium, cryogenics.

  13. Recent progress in the modelling of helium and tritium behaviour in irradiated beryllium pebbles

    Rabaglino, E.; Ronchi, C.; Cardella, A.

    2003-01-01

    One of the key issues of the European Helium Cooled Pebble Bed blanket is the behaviour under irradiation of beryllium pebbles, which have the function of neutron multiplier. An intense production of helium occurs in-pile, as well as a non negligible generation of tritium. Helium bubbles induce swelling and a high tritium inventory is a safety issue. Extensive studies for a better understanding, characterisation and modelling of the behaviour of helium and tritium in irradiated beryllium pebbles are being carried out, with the final aim to enable a reliable prediction of gas release and swelling in the full range of operating and accidental conditions of a Fusion Power Reactor. The general strategy consists in integrating studies on macroscopic phenomena (gas release) with the characterisation of corresponding microscopic diffusion phenomena (bubble kinetics) and the assessment of some fundamental diffusion parameter for the models (gas atomic diffusion coefficients). The present work gives a summary of the latest achievements in this context. By an inverse analysis of experimental out-of-pile gas release from weakly irradiated pebbles, coupled to the study of the characteristics of bubble population, it has been possible to assess the thermal diffusion coefficients of helium and tritium in and to improve and validate the classical model of gas precipitation into bubbles inside the grain. The improvement of the description of gas atomic diffusion and precipitation is the first step to enable a more reliable prediction of gas release

  14. Manufacturing pre-qualification of a Short Breeder Unit mockup (SHOBU) as part of the roadmap toward the out-of-pile validation of a full scale Helium Cooled Pebble Bed Breeder Unit

    Hernández, Francisco A.; Rey, Jorg; Neuberger, Heiko; Krasnorutskyi, Sergii; Niewöhner, Reinhard; Felde, Alexander

    2015-01-01

    Highlights: • A relevant mockup of a HCPB Breeder Unit for ITER (SHOBU) has been manufactured. • The manufacturing technologies used in SHOBU and its assembly sequence are reported. • Preliminary qualification of the welds has been successfully done after codes. • Future work foreseen to manufacture a feasibility mockup according to RCC-MRx code. - Abstract: The key components of the Helium Cooled Pebble Bed Test Blanket Module (HCPB TBM) in ITER are the Breeder Units (BU). These are the responsible for the tritium breeding and part of the heat extraction in the HCPB TBM. After a detailed design and engineering phase performed during the last years in the Karlsruhe Institute of Technology (KIT), a reference model for the manufacturing of a HCPB BU mock-up has been obtained. The mid-term is the out-of-pile qualification of the thermal and thermo-mechanical performance of a full-scale HCPB BU mock-up in a dedicated helium loop. Several key manufacturing technologies have been developed for the fabrication of the HCPB BU. In order to pre-qualify these techniques, a Short Breeder Unit mock-up (SHOBU) is under construction and to be tested. This paper aims at describing the relevance of SHOBU with a full-scale HCPB BU, the constitutive parts of SHOBU, the manufacturing and joining technologies involved, the assembly sequence (taking into consideration functional steps like its filling with Li_4SiO_4 pebbles or its assembly in the HCPB TBM) and the welding procedures studied. The paper concludes with a description of the required pre-qualification tests performed to SHOBU, i.e. pressure and leak tightness tests, according to the standards.

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

    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

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

    Peterson, Per; Greenspan, Ehud

    2015-01-01

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

  17. Helium cooling of fusion reactors

    Wong, C.P.C.; Baxi, C.; Bourque, R.; Dahms, C.; Inamati, S.; Ryder, R.; Sager, G.; Schleicher, R.

    1994-01-01

    On the basis of worldwide design experience and in coordination with the evolution of the International Thermonuclear Experimental Reactor (ITER) program, the application of helium as a coolant for fusion appears to be at the verge of a transition from conceptual design to engineering development. This paper presents a review of the use of helium as the coolant for fusion reactor blanket and divertor designs. The concept of a high-pressure helium cooling radial plate design was studied for both ITER and PULSAR. These designs can resolve many engineering issues, and can help with reaching the goals of low activation and high performance designs. The combination of helium cooling, advanced low-activation materials, and gas turbine technology may permit high thermal efficiency and reduced costs, resulting in the environmental advantages and competitive economics required to make fusion a 21st century power source. ((orig.))

  18. Tritium and helium release from beryllium pebbles neutron-irradiated up to 230appm tritium and 3000appm helium

    Vladimir Chakin

    2016-12-01

    Full Text Available Study of tritium and helium release from beryllium pebbles with diameters of 0.5 and 1mm after high-dose neutron irradiation at temperatures of 686–968K was performed. The release rate always has a single peak, and the peak temperatures at heating rates of 0.017K/s and 0.117K/s lie in the range of 1100–1350K for both tritium and helium release. The total tritium release from 1mm pebbles decreases considerably by increasing the irradiation temperature. The total tritium release from 0.5mm pebbles is less than that from 1mm pebbles and remains constant regardless of the irradiation temperature. At high irradiation temperatures, open channels are formed which contribute to the enhanced tritium release.

  19. Ceramic BOT type blanket with poloidal helium cooling

    Cardella, A.; Daenenr, W.; Iseli, M.; Ferrari, M.; Gallina, M.; Rado, V.; Simbolotti, G.; Violante, V.

    1989-01-01

    This paper briefly describes the work done and results achieved over the past two years on the ceramic breeder BOT blanket with poloidal helium cooling. A conclusive remark on the brick/plate option described previously is followed by short descriptions of the low and high performance pebble bed options elaborated as alternatives for both NET and DEMO. The results show, togethre with those about the poloidal cooling of the First Wall, good prospects for this blanket type provided that the questions connected wiht an extensive use of beryllium find a satisfactor answer. (author). 5 refs.; 7 figs.; 1 tab

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

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

    2016-07-15

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

  1. Helium-cooled nuclear reactor

    Longton, P.B.; Cowen, H.C.

    1975-01-01

    In helium cooled HTR's there is a by-pass circuit for cleaning purposes in addition to the main cooling circuit. This is to remove such impurities as hydrogen, methane, carbon monoxide and water from the coolant. In this system, part of the coolant successively flows first through an oxidation bed of copper oxide and an absorption bed of silica gel, then through activated charcoal or a molecular sieve. The hydrogen and carbon monoxide impurities are absorbed and the dry gas is returned to the main cooling circuit. To lower the hydrogen/water ratio without increasing the hydrogen fraction in the main cooling circuit, some of the hydrogen fraction converted into water is added to the cooling circuit. This is done, inter alia, by bypassing the water produced in the oxidation bed before it enters the absorption bed. The rest of the by-pass circuit, however, also includes an absorption bed with a molecular sieve. This absorbs the oxidized carbon monoxide fraction. In this way, such side effects as the formation of additional methane, carburization of the materials of the by-pass circuit or loss of graphite are avoided. (DG/RF) [de

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

    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.

  3. Pore Scale Thermal Hydraulics Investigations of Molten Salt Cooled Pebble Bed High Temperature Reactor with BCC and FCC Configurations

    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.

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

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

    2015-11-15

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

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

    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.

  6. Evaluation of helium cooling for fusion divertors

    Baxi, C.B.

    1993-09-01

    The divertors of future fusion reactors will have a power throughput of several hundred MW. The peak heat flux on the diverter surface is estimated to be 5 to 15 MW/m 2 at an average heat flux of 2 MW/m 2 . The divertors have a requirement of both minimum temperature (100 degrees C) and maximum temperature. The minimum temperature is dictated by the requirement to reduce the absorption of plasma, and the maximum temperature is determined by the thermo-mechanical properties of the plasma facing materials. Coolants that have been considered for fusion reactors are water, liquid metals and helium. Helium cooling has been shown to be very attractive from safety and other considerations. Helium is chemically and neutronically inert and is suitable for power conversion. The challenges associated with helium cooling are: (1) Manifold sizes; (2) Pumping power; and (3) Leak prevention. In this paper the first two of the above design issues are addressed. A variety of heat transfer enhancement techniques are considered to demonstrate that the manifold sizes and the pumping power can be reduced to acceptable levels. A helium-cooled diverter module was designed and fabricated by GA for steady-state heat flux of 10 MW/m 2 . This module was recently tested at Sandia National Laboratories. At an inlet pressure of 4 MPa, the module was tested at a steady-state heat flux of 10 MW/m 2 . The pumping power required was less than 1% of the power removed. These results verified the design prediction

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

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

    2015-01-15

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

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

    Gokhan Yesilyurt; Hassan, Y.A.

    2003-01-01

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

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

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

    2008-01-01

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

  10. Consideration of emergency source terms for pebble-bed high temperature gas-cooled reactor

    Tao, Liu; Jun, Zhao; Jiejuan, Tong; Jianzhu, Cao

    2009-01-01

    Being the last barrier in the nuclear power plant defense-in-depth strategy, emergency planning (EP) is an integrated project. One of the key elements in this process is emergency source terms selection. Emergency Source terms for light water reactor (LWR) nuclear power plant (NPP) have been introduced in many technical documents, and advanced NPP emergency planning is attracting attention recently. Commercial practices of advanced NPP are undergoing in the world, pebble-bed high-temperature gas-cooled reactor (HTGR) power plant is under construction in China which is considered as a representative of advanced NPP. The paper tries to find some pieces of suggestion from our investigation. The discussion of advanced NPP EP will be summarized first, and then the characteristics of pebble-bed HTGR relating to EP will be described. Finally, PSA insights on emergency source terms selection and current pebble-bed HTGR emergency source terms suggestions are proposed

  11. Cooling performance of helium-gas/water coolers in HENDEL

    Inagaki, Yoshiyuki; Takada, Shoji; Hayashi, Haruyoshi; Kobayashi, Toshiaki; Ohta, Yukimaru; Shimomura, Hiroaki; Miyamoto, Yoshiaki

    1994-01-01

    The helium engineering demonstration loop (HENDEL) has four helium-gas/water coolers where the cooling water flows in the tubes and helium gas on the shell side. Their cooling performance was studied using the operational data from 1982 to 1991. The heat transfer of helium gas on the shell was obtained for segmental and step-up baffle type coolers. Also, the change with operation time was investigated. The cooling performance was lowered by the graphite powder released from the graphite components for several thousand hours and thereafter recovered because the graphite powder from the components was reduced and the powder in the cooler shell was blown off during the operation. (orig.)

  12. Discussion on Design Transients of Pebble-bed High Temperature Gas-cooled Reactor

    Wang Yan; Li Fu; Zheng Yanhua

    2014-01-01

    In order to assure high quality for the components and their supports in the reactor coolant system, etc., some thermal-hydraulic transient conditions will be selected and researched for equipment design evaluation to satisfy the requirements ASME code, which are based on the conservative estimates of the magnitude and frequency of the temperature and pressure transients resulting from various operating conditions in the plant. In the mature design on pressurized water reactor, five conditions are considered. For the developing advanced pebble-bed high temperature gas-cooled reactor(HTGR), its design and operation has much difference with other reactors, so the transients of the pebble-bed high temperature gas-cooled reactor have distinctive characteristics. In this paper, the possible design transients of the pebble-bed HTGR will be discussed, and the frequency of design transients for equipment fatigue analysis and stress analysis due to cyclic stresses is also studied. The results will provide support for the design and construct of the pebble-bed HTGR. (author)

  13. Forced two phase helium cooling of large superconducting magnets

    Green, M.A.; Burns, W.A.; Taylor, J.D.

    1979-08-01

    A major problem shared by all large superconducting magnets is the cryogenic cooling system. Most large magnets are cooled by some variation of the helium bath. Helium bath cooling becomes more and more troublesome as the size of the magnet grows and as geometric constraints come into play. An alternative approach to cooling large magnet systems is the forced flow, two phase helium system. The advantages of two phase cooling in many magnet systems are shown. The design of a two phase helium system, with its control dewar, is presented. The paper discusses pressure drop of a two phase system, stability of a two phase system and the method of cool down of a two phase system. The results of experimental measurements at LBL are discussed. Included are the results of cool down and operation of superconducting solenoids

  14. Behavior of beryllium pebbles under irradiation

    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)

  15. Probabilistic safety assessment framework of pebble-bed modular high-temperature gas-cooled reactor

    Liu Tao; Tong Jiejuan; Zhao Jun; Cao Jianzhu; Zhang Liguo

    2009-01-01

    After an investigation of similar reactor type probabilistic safety assessment (PSA) framework, Pebble-bed Modular High-Temperature Gas-cooled Reactor (HTR-PM) PSA framework was presented in correlate with its own design characteristics. That is an integral framework which spreads through event sequence structure with initiating events at the beginning and source term categories in the end. The analysis shows that it is HTR-PM design feature that determines its PSA framework. (authors)

  16. An efficient continuous flow helium cooling unit for Moessbauer experiments

    Herbert, I.R.; Campbell, S.J.

    1976-01-01

    A Moessbauer continuous flow cooling unit for use with liquid helium over the temperature range 4.2 to 300K is described. The cooling unit can be used for either absorber or source studies in the horizontal plane and it is positioned directly on top of a helium storage vessel. The helium transfer line forms an integral part of the cooling unit and feeds directly into the storage vessel so that helium losses are kept to the minimum. The helium consumption is 0.12 l h -1 at 4.2 K decreasing to 0.055 l h -1 at 40 K. The unit is top loading and the exchange gas cooled samples can be changed easily and quickly. (author)

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

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

    2006-01-01

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

  18. Cold flow study of liquid cooled pebble bed reactor (LC-PBR) through radioisotope techniques

    Verma, Rupesh; Upadhyay, Rajesh K.; Pant, H.J.

    2017-01-01

    As the world's demand for energy continues to increase burning of coal, oil and natural gases continue to increase which will eventually cause build-up in emission of greenhouse gasses. To overcome this challenge worldwide effort is in progress to develop an economical, more efficient and safer nuclear power. Higher thermal efficiency and enhances safety feature of Generation IV liquid cooled pebble bed reactor (LC-PBR) makes it viable option to replace existing nuclear reactor. However, this reactor is still in research stage and need detailed study before commercialization. In current work, hydrodynamics of LC-PBR is studied by using radioisotope based techniques, radioactive particle tracking and gamma-ray densitometry. Pebble flow profile and distribution are measured for different operating conditions. Optimal operating parameters are identified for operating LC-PBR based on hydrodynamics. (author)

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

    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

  20. Failure initiation and propagation of Li4SiO4 pebbles in fusion blankets

    Zhao Shuo; Gan Yixiang; Kamlah, Marc

    2013-01-01

    Lithium orthosilicate (Li 4 SiO 4 ) pebbles are considered to be a candidate as solid tritium breeder in the helium cooled pebble bed (HCPB) blanket. These ceramic pebbles might be crushed during thermomechanical loading in the blanket. In this work, the failure initiation and propagation of pebbles in pebble beds is investigated using the discrete element method (DEM). Pebbles are simplified as mono-sized elastic spheres. Every pebble has a contact strength in terms of critical strain energy, which is derived from a validated strength model and crush test data for pebbles from a specific batch of Li 4 SiO 4 pebbles. Pebble beds are compressed uniaxially and triaxially in DEM simulations. When the strain energy absorbed by a pebble exceeds its critical energy it fails. The failure initiation is defined as a given small fraction of pebbles crushed. It is found that the load level for failure initiation can be very low. For example, if failure initiation is defined as soon as 0.02% of the pebbles have been crushed, the pressure required for uniaxial loading is about 2.5 MPa. Therefore, it is essential to study the influence of failure propagation on the macroscopic response of pebble beds. Thus a reduction ratio defined as the size ratio of a pebble before and after its failure is introduced. The macroscopic stress–strain relation is investigated with different reduction ratios. A typical stress plateau is found for a small reduction ratio.

  1. LOFA analyses for the water and helium cooled SEAFP reactors

    Sponton, L.; Sjoeberg, A.; Nordlinder, S.

    2001-01-01

    This study was performed in the frame of the European long-term fusion safety programme 1999 (SEAFP99). Loss of flow accidents (LOFA) have been studied for two cases, first for a helium cooled reactor with advanced dual-coolant (DUAL) blanket at 100% nominal power. The second case applies to a water-cooled reactor at 20% nominal power. Both transients were simulated with the code MELCOR 1.8.4. The results for the helium cooled reactor show that with a natural circulation flow of helium after the pump stops, the first wall temperature will stay below the temperature for excepted failure of the construction material. For the water cooled reactor, the results show that the pressurizer set point for its liquid volumetric inventory is reached before the plasma facing components attain a critical temperature. The pressurizer set point will induce a plasma shutdown

  2. The cryogenic helium cooling system for the Tokamak physics experiment

    Felker, B.; Slack, D.S.; Wendland, C.R.

    1995-01-01

    The Tokamak Physics Experiment (TPX) will use supercritical helium to cool all the magnets and supply helium to the Vacuum cryopumping subsystem. The heat loads will come from the standard steady state conduction and thermal radiation sources and from the pulsed loads of the nuclear and eddy currents caused by the Central Solenoid Coils and the plasma positioning coils. The operations of the TPX will begin with pulses of up to 1000 seconds in duration every 75 minutes. The helium system utilizes a pulse load leveling scheme to buffer out the effects of the pulse load and maintain a constant cryogenic plant operation. The pulse load leveling scheme utilizes the thermal mass of liquid and gaseous helium stored in a remote dewar to absorb the pulses of the tokamak loads. The mass of the stored helium will buffer out the temperature pulses allowing 5 K helium to be delivered to the magnets throughout the length of the pulse. The temperature of the dewar will remain below 5 K with all the energy of the pulse absorbed. This paper will present the details of the heat load sources, of the pulse load leveling scheme operations, a partial helium schematic, dewar temperature as a function of time, the heat load sources as a function of time and the helium temperature as a function of length along the various components that will be cooled

  3. A methodology for accident analysis of fusion breeder blankets and its application to helium-cooled lead–lithium blanket

    Panayotov, Dobromir; Poitevin, Yves; Grief, Andrew; Trow, Martin; Dillistone, Michael

    2016-01-01

    'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. Furthermore, the methodology phases are illustrated in the paper by its application to the EU HCLL TBS using both MELCOR and RELAP5 codes.

  4. Evaluation of US demo helium-cooled blanket options

    Wong, C.P.C.; McQuillan, B.W.; Schleicher, R.W.

    1995-10-01

    A He-V-Li blanket design was developed as a candidate for the U.S. fusion demonstration power plant. This paper presents an 18 MPa helium-cooled, lithium breeder, V-alloy design that can be coupled to the Brayton cycle with a gross efficiency of 46%. The critical issue of designing to high gas pressure and the compatibility between helium impurities and V-alloy are addressed

  5. Liquid helium cooling of the MFTF superconducting magnets

    VanSant, J.H.; Zbasnik, J.P.

    1986-09-01

    During acceptance testing of the Mirror Fusion Test Facility (MFTF), we measured these tests: liquid helium heat loads and flow rates in selected magnets. We used the data from these tests to estimate helium vapor quality in the magnets so that we could determine if adequate conductor cooling conditions had occurred. We compared the measured quality and flow with estimates from a theoretical model developed for the MFTF magnets. The comparison is reasonably good, considering influences that can greatly affect these values. This paper describes the methods employed in making the measurements and developing the theoretical estimates. It also describes the helium system that maintained the magnets at required operating conditions

  6. Cooling by mixing of helium isotopes

    Hansen, O.P.; Olsen, M.; Rasmussen, F.B.

    1975-01-01

    The principles of the helium dilution refrigerator are outlined. The lowest temperature attained with a continuously operated dilution refrigerator was about 10 mK, and 5 mK for a limited period when the supply of concentrated 3 He to the mixing chamber was interrupted. (R.S.)

  7. The evolution of US helium-cooled blankets

    Wong, C.P.C.; Schultz, K.R.; Cheng, E.T.

    1991-01-01

    This paper reviews and compares four helium-cooled fusion reactor blanket designs. These designs represent generic configurations of using helium to cool fusion reactor blankets that were studied over the past 20 years in the United States of America (US). These configurations are the pressurized module design, the pressurized tube design, the solid particulate and gas mixture design, and the nested shell design. Among these four designs, the nested shell design, which was invented for the ARIES study, is the simplest in configuration and has the least number of critical issues. Both metallic and ceramic-composite structural materials can be used for this design. It is believed that the nested shell design can be the most suitable blanket configuration for helium-cooled fusion power and experimental reactors. (orig.)

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

    Yang Xingtuan; Hu Wenping; Jiang Shengyao

    2009-01-01

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

  9. Supercritical Helium Cooling of the LHC Beam Screens

    Hatchadourian, E; Tavian, L

    1998-01-01

    The cold mass of the LHC superconducting magnets, operating in pressurised superfluid helium at 1.9 K, must be shielded from the dynamic heat loads induced by the circulating particle beams, by means of beam screens maintained at higher temperature. The beam screens are cooled between 5 and 20 K by forced flow of weakly supercritical helium, a solution which avoids two-phase flow in the long, narr ow cooling channels, but still presents a potential risk of thermohydraulic instabilities. This problem has been studied by theoretical modelling and experiments performed on a full-scale dedicated te st loop.

  10. Status of helium-cooled nuclear power systems. [Development potential

    Melese-d' Hospital, G.; Simnad, M

    1977-09-01

    Helium-cooled nuclear power systems offer a great potential for electricity generation when their long-term economic, environmental, conservation and energy self-sufficiency features are examined. The high-temperature gas-cooled reactor (HTGR) has the unique capability of providing high-temperature steam for electric power and process heat uses and/or high-temperature heat for endothermic chemical reactions. A variation of the standard steam cycle HTGR is one in which the helium coolant flows directly from the core to one or more closed cycle gas turbines. The effective use of nuclear fuel resources for electric power and nuclear process heat will be greatly enhanced by the gas-cooled fast breeder reactor (GCFR) currently being developed. A GCFR using thorium in the radial blanket could generate sufficient U-233 to supply the fuel for three HTGRs, or enough plutonium from a depleted uranium blanket to fuel a breeder economy expanding at about 10% per year. The feasibility of utilizing helium to cool a fusion reactor is also discussed. The status of helium-cooled nuclear energy systems is summarized as a basis for assessing their prospects. 50 references.

  11. On natural circulation in High Temperature Gas-Cooled Reactors and pebble bed reactors for different flow regimes and various coolant gases

    Melesed'Hospital, G.

    1983-01-01

    The use of CO 2 or N 2 (heavy gas) instead of helium during natural circulation leads to improved performance in both High Temperature Gas-Cooled Reactors (HTGR) and in Pebble Bed Reactors (PBR). For instance, the coolant temperature rise corresponding to a coolant pressure level and a rate of afterheat removal could be only 18% with CO 2 as compared to He, for laminar flow in HTGR; this value would be 40% in PBR. There is less difference between HTGR and PBR for turbulent flows; CO 2 is found to be always better than N 2 . These types of results derived from relationships between coolant properties, coolant flow, temperature rise, pressure, afterheat levels and core geometry, are obtained for HTGR and PBR for various flow regimes, both within the core and in the primary loop

  12. Gas cooled high temperature reactor with a heap of pebble shaped fuel elements and absorber rods which can be driven directly into the heap of pebbles

    Elter, C.; Schmitt, H.; Schoening, J.; Weicht, U.

    1980-01-01

    The absorber rod for the graphite moderated, helium cooled reactor is cylindrical and has a tip in the shape of the frustrum of a cone. It consists of three coaxially arranged sleeve tubes made of steel, the inner and centre sleeve tubes surrounding the absorber part (B4C) so as to be gastight. The inner sleeve tube represents the supporting tube and is cooled by cold gas, as is the annular gap between the centre and outer sleeve tube. (RW) [de

  13. Radiolytic reactions in the coolant of helium cooled reactors

    Tingey, G.L.; Morgan, W.C.

    1975-01-01

    The success of helium cooled reactors is dependent upon the ability to prevent significant reaction between the coolant and the other components in the reactor primary circuit. Since the thermal reaction of graphite with oxidizing gases is rapid at temperatures of interest, the thermal reactions are limited primarily by the concentration of impurity gases in the helium coolant. On the other hand, the rates of radiolytic reactions in helium are shown to be independent of reactive gas concentration until that concentration reaches a very low level. Calculated steady-state concentrations of reactive species in the reactor coolant and core burnoff rates are presented for current U. S. designed, helium cooled reactors. Since precise base data are not currently available for radiolytic rates of some reactions and thermal reaction rate data are often variable, the accuracy of the predicted gas composition is being compared with the actual gas compositions measured during startup tests of the Fort Saint Vrain high temperature gas-cooled reactor. The current status of these confirmatory tests is discussed. 12 references

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

    Wei Renjie

    1998-01-01

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

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

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

    2016-01-01

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

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

    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.

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

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

    1981-01-01

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

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

    Yuan Tao; Feng Kaiming; Chen Zhi; Wang Xiaoyu

    2007-01-01

    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 (Li 4 SiO 4 ) 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. Gas Turbine High Temperature Gas (Helium) Reactor Using Pebble Bed Fuel Derived from Spent Fuel

    Cole, Quentin

    2013-01-01

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

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

    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.

  1. Numerical investigation of the flow at the pebble bed of the high temperature gas cooled reactors

    Costa, Franklin C.; Navarro, Moyses A.; Santos, Andre A.C.

    2011-01-01

    This paper presents a numerical investigation of the thermal and fluid dynamics among the fuel spheres and the cooling fluid, appearing in the core of pebble bed reactor (PBR-Peeble Bed Reactor) using the CFD-Computational Fluid Dynamics CFX 13.0. The paper presents the two analysis results. In the first phase it was considered two heat transfer models for the fuel spheres. In a model it was established volumetric load generation, with thermal conduction for both the fuel and coating. The other model prescribes a heat flux at the sphere surfaces. In this analysis, it was proceed two simulation in the two sphere arrangements, one considering the spheres in contact, and the other with 2 mm spacing between them. At the second analysis it was evaluated the sphere arrangement influence on the thermal and fluid dynamic behavior of the bed. The four simulations present differences in the flow and in the surface and maximum temperature profiles of the coating.(author)

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

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

    2014-01-01

    The presence of water on the tube-side of the steam generators in high temperature gas-cooled reactors (HTGRs) with indirect cycle layouts presents a possibility for a penetration of neutron moderating steam into the core, which may cause a power excursion. This article presents results on the effect of water ingress into the core of the two South African Pebble Bed Modular Reactor design concepts, i.e. the PBMR-200 MW th and the PBMR-400 MW th developed by PBMR SOC Ltd. The VSOP 99/05 suite of codes was used for the simulation of this event. Partial steam vapour pressures were added in stages into the primary circuit in order to investigate the effect of water ingress on reactivity, power profiles and thermal neutron flux profiles. The effects of water ingress into the core are explained by increased neutron moderation, due to the addition of 1 H, which leads to a decrease in resonance capture by 238 U and therefore an increase in the multiplication factor. The more effective moderation of neutrons by definition reduces the fast neutron flux and increases the thermal flux in the core, i.e. leads to a softer spectrum. The more effective moderation also increases the average increase in lethargy between collisions of a neutron with successive fuel kernels, which reduces the probability for neutron capture in the radiative capture resonances of 238 U. The resulting higher resonance escape probability also increases the thermal flux in the core. The softening of the neutron spectrum leads to an increased effective microscopic fission cross section in the fissile isotopes and thus to increased neutron absorption for fission, which reduces the remaining number of neutrons that can diffuse into the reflectors. Therefore water ingress into the core leads to a reduced thermal neutron flux in the reflectors. The power density spatial distribution behaved similarly to the thermal neutron flux in the core. Analysis of possible mechanisms was conducted. The results show that

  3. Heat Transfer in Pebble-Bed Nuclear Reactor Cores Cooled by Fluoride Salts

    Huddar, Lakshana Ravindranath

    With electricity demand predicted to rise by more than 50% within the next 20 years and a burgeoning world population requiring reliable emissions-free base-load electricity, can we design advanced nuclear reactors to help meet this challenge? At the University of California, Berkeley (UCB) Fluoride-salt-cooled High Temperature Reactors (FHR) are currently being investigated. FHRs are designed with better safety and economic characteristics than conventional light water reactors (LWR) currently in operation. These reactors operate at high temperature and low pressure making them more efficient and safer than LWRs. The pebble-bed FHR (PB-FHR) variant includes an annular nuclear reactor core that is filled with randomly packed pebble fuel. It is crucial to characterize the heat transfer within this unique geometry as this informs the safety limits of the reactor. The work presented in this dissertation focused on furthering the understanding of heat transfer in pebble-bed nuclear reactor cores using fluoride salts as a coolant. This was done through experimental, analytical and computational techniques. A complex nuclear system with a coolant that has never previously been in commercial use requires experimental data that can directly inform aspects of its design. It is important to isolate heat transfer phenomena in order to understand the underlying physics in the context of the PB-FHR, as well as to make decisions about further experimental work that needs to be done in support of developing the PB-FHR. Certain organic oils can simulate the heat transfer behaviour of the fluoride salt if relevant non-dimensional parameters are matched. The advantage of this method is that experiments can be done at a much lower temperature and at a smaller geometric scale compared to FHRs, thereby lowering costs. In this dissertation, experiments were designed and performed to collect data demonstrating similitude. The limitations of these experiments were also elucidated by

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

    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)

  5. Status of the in-pile test of HCPB pebble-bed assemblies in the HFR Petten

    Laan, J.G. van der; Fokkens, J.H.; Hofmans, H.E.; Jong, M.; Magielsen, A.J.; Pijlgroms, B.J.; Stijkel, M.P. [NRG, Petten (Netherlands); Conrad, R. [JRC, Inst. for Energy, Petten (Netherlands); Malang, S.; Reimann, J. [FZK, Karlsruhe (Germany); Roux, N. [CEA Saclay (France)

    2002-06-01

    In the framework of developing the helium cooled pebble-bed (HCPB) blanket an irradiation test of pebble-bed assemblies is prepared at the HFR Petten. The test objective is to concentrate on the effect of neutron irradiation on the thermal-mechanical behaviour of the HCPB breeder pebble-bed at DEMO representative levels of temperature and defined thermal-mechanical loads. The basic test elements are EUROFER-97 cylinders with a horizontal bed of ceramic breeder pebbles sandwiched between two beryllium beds. The pebble beds are separated by EUROFER-97 steel plates. The heat flow is managed such as to have a radial temperature distribution in the ceramic breeder pebble-bed as flat as reasonably possible. The paper reports on the project status, and presents the results of pre-tests, material characteristics, the manufacturing of the pebble-bed assemblies, and the nuclear and thermo-mechanical loading parameters. (orig.)

  6. Preliminary Overview of a Helium Cooling System for the Secondary Helium Loop in VHTR-based SI Hydrogen Production Facilities

    Shin, Youngjoon; Cho, Mintaek; Kim, Dahee; Lee, Taehoon; Lee, Kiyoung; Kim, Yongwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Nuclear hydrogen production facilities consist of a very high temperature gas-cooled nuclear reactor (VHTR) system, intermediate heat exchanger (IHX) system, and a sulfur-iodine (SI) thermochemical process. This study focuses on the coupling system between the IHX system and SI thermochemical process. To prevent the propagation of the thermal disturbance owing to the abnormal operation of the SI process components from the IHX system to the VHTR system, a helium cooling system for the secondary helium of the IHX is required. In this paper, the helium cooling system has been studied. The temperature fluctuation of the secondary helium owing to the abnormal operation of the SI process was then calculated based on the proposed coupling system model. Finally, the preliminary conceptual design of the helium cooling system with a steam generator and forced-draft air-cooled heat exchanger to mitigate the thermal disturbance has been carried out. A conceptual flow diagram of a helium cooling system between the IHX and SI thermochemical processes in VHTR-based SI hydrogen production facilities has been proposed. A helium cooling system for the secondary helium of the IHX in this flow diagram prevents the propagation of the thermal disturbance from the IHX system to the VHTR system, owing to the abnormal operation of the SI process components. As a result of a dynamic simulation to anticipate the fluctuations of the secondary helium temperature owing to the abnormal operation of the SI process components with a hydrogen production rate of 60 mol·H{sub 2}/s, it is recommended that the maximum helium cooling capacity to recover the normal operation temperature of 450 .deg. C is 31,933.4 kJ/s. To satisfy this helium cooling capacity, a U-type steam generator, which has a heat transfer area of 12 m{sup 2}, and a forced-draft air-cooled condenser, which has a heat transfer area of 12,388.67 m{sup 2}, are required for the secondary helium cooling system.

  7. Surface coating of graphite pebbles for Korean HCCR TBM

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

    2014-10-15

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

  8. Surface coating of graphite pebbles for Korean HCCR TBM

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

    2014-01-01

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

  9. Test-element assembly and loading parameters for the in-pile test of HCPB ceramic pebble beds

    Laan, J.G. van der E-mail: vanderlaan@nrg-nl.com; Boccaccini, L.V.; Conrad, R.; Fokkens, J.H.; Jong, M.; Magielsen, A.J.; Pijlgroms, B.J.; Reimann, J.; Stijkel, M.P.; Malang, S

    2002-11-01

    In the framework of developing the helium cooled pebble-bed (HCPB) blanket an irradiation test of pebble-bed assemblies is prepared at the HFR Petten. The test objective is to concentrate on the effect of neutron irradiation on the thermal-mechanical behaviour of the HCPB breeder pebble-bed at DEMO representative levels of temperature and defined thermal-mechanical loads. The paper reports on the project status, and presents the results of pre-tests, material characteristics, the manufacturing of the pebble-bed assemblies, and the nuclear and thermo-mechanical loading parameters.

  10. [Fluid dynamics of supercritical helium within internally cooled cabled superconductors

    Van Sciver, S.W.

    1995-01-01

    The Applied Superconductivity Center of the University of Wisconsin-Madison proposes to conduct research on low temperature helium fluid dynamics as it applies to the cooling of internally cooled cabled superconductors (ICCS). Such conductors are used in fusion reactor designs including most of the coils in ITER. The proposed work is primarily experimental involving measurements of transient and steady state pressure drop in a variety of conductor configurations. Both model and prototype conductors for actual magnet designs will be investigated. The primary goal will be to measure and model the friction factor for these complex geometries. In addition, an effort will be made to study transient processes such as heat transfer and fluid expulsion associated with quench conditions

  11. Helium-Cooled Refractory Alloys First Wall and Blanket Evaluation

    Wong, C.P.C.; Nygren, R.E.; Baxi, C.B.; Fogarty, P.; Ghoniem, N.; Khater, H.; McCarthy, K.; Merrill, B.; Nelson, B.; Reis, E.E.; Sharafat, S.; Schleicher, R.; Sze, D.K.; Ulrickson, M.; Willms, S.; Youssef, M.; Zinkel, S.

    1999-01-01

    Under the APEX program the He-cooled system design task is to evaluate and recommend high power density refractory alloy first wall and blanket designs and to recommend and initiate tests to address critical issues. We completed the preliminary design of a helium-cooled, W-5Re alloy, lithium breeder design and the results are reported in this paper. Many areas of the design were assessed, including material selection, helium impurity control, and mechanical, nuclear and thermal hydraulics design, and waste disposal, tritium and safety design. System study results show that at a closed cycle gas turbine (CCGT) gross thermal efficiency of 57.5%, a superconducting coil tokamak reactor, with an aspect ratio of 4, and an output power of 2 GWe, can be projected to have a cost of electricity at 54.6 mill/kWh. Critical issues were identified and we plan to continue the design on some of the critical issues during the next phase of the APEX design study

  12. Measurement of flow field in the pebble bed type high temperature gas-cooled reactor

    Lee, Sa Ya; Lee, Jae Young

    2008-01-01

    In this study, flow field measurement of the Pebble Bed Reactor(PBR) for the High Temperature Gascooled Reactor(HTGR) was performed. Large number of pebbles in the core of PBR provides complicated flow channel. Due to the complicated geometries, numerical analysis has been intensively made rather than experimental observation. However, the justification of computational simulation by the experimental study is crucial to develop solid analysis of design method. In the present study, a wind tunnel installed with pebbles stacked was constructed and equipped with the Particle Image Velocimetry(PIV). We designed the system scaled up to realize the room temperature condition according to the similarity. The PIV observation gave us stagnation points, low speed region so that the suspected high temperature region can be identified. With the further supplementary experimental works, the present system may produce valuable data to justify the Computational Fluid Dynamics(CFD) simulation method

  13. Superconducting cable cooling system by helium gas at two pressures

    Dean, J.W.

    1977-01-01

    Thermally contacting, oppositely streaming, cryogenic fluid streams in the same enclosure in a closed cycle changes the fluid from a cool high pressure helium gas to a cooler reduced pressure helium gas in an expander so as to be at different temperature ranges and pressures respectively in go and return legs that are in thermal contact with each other and in thermal contact with a longitudinally extending superconducting transmission line enclosed in the same cable enclosure that insulates the line from the ambient at a temperature T 1 . By first circulating the fluid from a refrigerator at one end of the line as a cool gas at a temperature range T 2 to T 3 in the go leg, then circulating the gas through an expander at the other end of the line where the gas becomes a cooler gas at a reduced pressure and at a reduced temperature T 4 and finally by circulating the cooler gas back again to the refrigerator in a return leg at a temperature range T 4 to T 5 , while in thermal contact with the gas in the go leg, and in the same enclosure therewith for compression into a higher pressure gas at T 2 in a closed cycle, where T 2 greater than T 3 and T 5 greater than T 4 , the fluid leaves the enclosure in the go leg as a gas at its coldest point in the go leg, and the temperature distribution is such that the line temperature decreases along its length from the refrigerator due to the cooling from the gas in the return leg

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

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

    2010-01-01

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

  15. Study on fundamental features of helium turbomachine for high temperature gas-cooled reactor

    Wang Jie; Gu Yihua

    2004-01-01

    The High temperature gas-cooled reactor (HTGR) coupled with helium turbine cycle is considered as one of the leading candidates for future nuclear power plants. The HTGR helium turbine cycle was analyzed and optimized. Then the focal point of investigation was concentrated on the fundamental thermodynamic and aerodynamic features of helium turbomachine. As a result, a helium turbomachine is different from a general combustion gas turbine in two main design features, that is a helium turbomachine has more blade stages and shorter blade length, which are caused by the helium property and the high pressure of a closed cycle, respectively. (authors)

  16. The early history of high-temperature helium gas-cooled nuclear power reactors

    Simnad, M.T.; California Univ., San Diego, La Jolla, CA

    1991-01-01

    The original concepts in the proposals for high-temperature helium gas-cooled power reactors by Farrington Daniels, during the decade 1944-1955, are summarized. The early research on the development of the helium gas-cooled power reactors is reviewed, and the operational experiences with the first generation of HTGRs are discussed. (author)

  17. Cryogenic recovery analysis of forced flow supercritical helium cooled superconductors

    Lee, A.Y.

    1977-08-01

    A coupled heat conduction and fluid flow method of solution was presented for cryogenic stability analysis of cabled composite superconductors of large scale magnetic coils. The coils are cooled by forced flow supercritical helium in parallel flow channels. The coolant flow reduction in one of the channels during the spontaneous recovery transient, after the conductor undergoes a transition from superconducting to resistive, necessitates a parallel channel analysis. A way to simulate the parallel channel analysis is described to calculate the initial channel inlet flow rate required for recovery after a given amount of heat is deposited. The recovery capability of a NbTi plus copper composite superconductor design is analyzed and the results presented. If the hydraulics of the coolant flow is neglected in the recovery analysis, the recovery capability of the superconductor will be over-predicted

  18. 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

    Norajitra, P.

    1995-06-01

    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 Li 4 SiO 4 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.) [de

  19. Optimal thermal-hydraulic performance for helium-cooled divertors

    Izenson, M.G.; Martin, J.L.

    1996-01-01

    Normal flow heat exchanger (NFHX) technology offers the potential for cooling divertor panels with reduced pressure drops (<0.5% Δp/p), reduced pumping power (<0.75% pumping/thermal power), and smaller duct sizes than conventional helium heat exchangers. Furthermore, the NFHX can easily be fabricated in the large sizes required for divertors in large tokamaks. Recent experimental and computational results from a program to develop NFHX technology for divertor coolings using porous metal heat transfer media are described. We have tested the thermal and flow characteristics of porous metals and identified the optimal heat transfer material for the divertor heat exchanger. Methods have been developed to create highly conductive thermal bonds between the porous material and a solid substrate. Computational fluid dynamics calculations of flow and heat transfer in the porous metal layer have shown the capability of high thermal effectiveness. An 18-kW NFHX, designed to meet specifications for the international Thermonuclear Experimental Reactor divertor, has been fabricated and tested for thermal and flow performance. Preliminary results confirm design and fabrication methods. 11 refs., 12 figs., 1 tab

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

    Daigo Tsuru; Mikio Enoeda; Masato Akiba

    2006-01-01

    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

  1. First results of the post-irradiation examination of the Ceramic Breeder materials from the Pebble Bed Assemblies Irradiation for the HCPB Blanket concept

    Hegeman, J.; Magielsen, A.J.; Peeters, M.; Stijkel, M.P.; Fokkens, J.H.; Laan, J.G. van der

    2006-01-01

    In the framework of developing the European Helium Cooled Pebble-Bed (HCPB) blanket an irradiation test of pebble-bed assemblies is performed in the HFR Petten. The experiment is focused on the thermo-mechanical behavior of the HCPB type breeder pebble-bed at DEMO representative levels of temperature and defined thermal-mechanical loads. To achieve representative conditions a section of the HCPB is simulated by EUROFER-97 cylinders with a horizontal bed of ceramic breeder pebbles sandwiched between two beryllium beds. Floating Eurofer-97 steel plates separate the pebble-beds. The structural integrity of the ceramic breeder materials is an issue for the design of the Helium Cooled Pebble Bed concept. Therefore the objective of the post irradiation examination is to study deformation of pebbles and the pebble beds and to investigate the microstructure of the ceramic pebbles from the Pebble Bed Assemblies. This paper concentrates on the Post Irradiation Examination (PIE) of the four ceramic pebble beds that have been irradiated in the Pebble Bed Assembly experiment for the HCPB blanket concept. Two assemblies with Li 4 SiO 4 pebble-beds are operated at different maximum temperatures of approximately 600 o C and 800 o C. Post irradiation computational analysis has shown that both have different creep deformation. Two other assemblies have been loaded with a ceramic breeder bed of two types of Li 2 TiO 3 beds having different sintering temperatures and consequently different creep behavior. The irradiation maximum temperature of the Li 2 TiO 3 was 800 o C. To support the first PIE result, the post irradiation thermal analysis will be discussed because thermal gradients have influence on the pebble-bed thermo-mechanical behavior and as a result it may have impact on the structural integrity of the ceramic breeder materials. (author)

  2. Evaluation of radiation heat transfer in porous medial: Application for a pebble bed modular reactor cooled by CO2 gas

    Sidi-Ali Kamel

    2013-01-01

    Full Text Available This work analyses the contribution of radiation heat transfer in the cooling of a pebble bed modular reactor. The mathematical model, developed for a porous medium, is based on a set of equations applied to an annular geometry. Previous major works dealing with the subject have considered the forced convection mode and often did not take into account the radiation heat transfer. In this work, only free convection and radiation heat transfer are considered. This can occur during the removal of residual heat after shutdown or during an emergency situation. In order to derive the governing equations of radiation heat transfer, a steady-state in an isotropic and emissive porous medium (CO2 is considered. The obtained system of equations is written in a dimensionless form and then solved. In order to evaluate the effect of radiation heat transfer on the total heat removed, an analytical method for solving the system of equations is used. The results allow quantifying both radiation and free convection heat transfer. For the studied situation, they show that, in a pebble bed modular reactor, more than 70% of heat is removed by radiation heat transfer when CO2 is used as the coolant gas.

  3. Manufacturing and joining technologies for helium cooled divertors

    Aktaa, J.; Basuki, W.W.; Weber, T.; Norajitra, P.; Krauss, W.; Konys, J.

    2014-01-01

    Highlights: • The manufacturing and joining technologies developed at KIT for helium cooled divertors are reviewed and critically discussed. • Various technologies have been pursued and further developed aiming divertor components with very high quality and sufficient reliability. • Very promising routes have been found for which however still R and D works are necessary. • Technologies developed are also useful for other divertor and even blanket concepts, particularly those with tungsten armor. - Abstract: In the helium cooled (HC) divertor, developed at KIT for a fusion power plant, tungsten has been selected as armor as well as structural material due to its crucial properties: high melting point, very low sputtering yield, good thermal conductivity, high temperature strength, low thermal expansion and low activation. Thereby the armor tungsten is attached to the structural tungsten by thermally conductive joint. Due to the brittleness of tungsten at low temperatures its use as structural material is limited to the high temperature part of the component and a structural joint to the reduced activation ferritic martensitic steel EUROFER97 is foreseen. Hence, to realize the selected hybrid material concept reliable tungsten–steel and tungsten–tungsten joints have been developed and will be reported in this paper. In addition, the modular design of the HC divertor requires tungsten armor tiles and tungsten structural thimbles to be manufactured in high numbers with very high quality. Due to the high strength and low temperature brittleness of tungsten special manufacturing techniques need to be developed for the production of parts with no cavities inside and/or surface flaws. The main achievement in developing the respective manufacturing technologies will be presented and discussed. To achieve the objectives mentioned above various manufacturing and joining technologies are pursued. Their later applicability depends on the level of development

  4. High temperature helium-cooled fast reactor (HTHFR)

    Karam, R.A.; Blaylock, Dwayne; Burgett, Eric; Mostafa Ghiaasiaan, S.; Hertel, Nolan

    2006-01-01

    Scoping calculations have been performed for a very high temperature (1000 o C) helium-cooled fast reactor involving two distinct options: (1) using graphite foam into which UC (12% enrichment) is embedded into a matrix comprising UC and graphite foam molded into hexagonal building blocks and encapsulated with a SiC shell covering all surfaces, and (2) using UC only (also 12% enrichment) molded into the same shape and size as the foam-UC matrix in option 1. Both options use the same basic hexagonal fuel matrix blocks to form the core and reflector. The reflector contains natural uranium only. Both options use 50 μm SiC as a containment shell for fission product retention within each hexagonal block. The calculations show that the option using foam (option 1) would produce a reactor that can operate continuously for at least 25 years without ever adding or removing any fuel from the reactor. The calculations show further that the UC only option (option 2) can operate continually for 50 years without ever adding or removing fuel from the reactor. Doppler and loss of coolant reactivity coefficients were calculated. The Doppler coefficient is negative and much larger than the loss of coolant coefficient, which was very small and positive. Additional progress on and development of the two concepts are continuing

  5. A review of helium gas turbine technology for high-temperature gas-cooled reactors

    No, Hee Cheon; Kim, Ji Hwan; Kim, Hyeun Min

    2007-01-01

    Current High-Temperature Gas-cooled Reactors (HTGRs) are based on a closed brayton cycle with helium gas as the working fluid. Thermodynamic performance of the axial-flow helium gas turbines is of critical concern as it considerably affects the overall cycle efficiency. Helium gas turbines pose some design challenges compared to steam or air turbomachinery because of the physical properties of helium and the uniqueness of the operating conditions at high pressure with low pressure ratio. This report present a review of the helium Brayton cycle experiences in Germany and in Japan. The design and availability of helium gas turbines for HTGR are also presented in this study. We have developed a new throughflow calculation code to calculate the design-point performance of helium gas turbines. Use of the method has been illustrated by applying it to the GTHTR300 reference

  6. Liquid helium-cooled MOSFET preamplifier for use with astronomical bolometer

    Goebel, J. H.

    1977-01-01

    A liquid helium-cooled p-channel enhancement mode MOSFET, the 3N167, is found to have sufficiently low noise for use as a preamplifier with helium-cooled bolometers that are used in infrared astronomy. Its characteristics at 300, 77, and 4.2 K are presented. It is also shown to have useful application with certain photoconductive and photovoltaic infrared detectors.

  7. Solutions for Liquid Nitrogen Pre-Cooling in Helium Refrigeration Cycles

    Wagner, U

    2000-01-01

    Pre-cooling of helium by means of liquid nitrogen is the oldest and one of the most common process features used in helium liquefiers and refrigerators. Its two principle tasks are to allow or increase the rate of pure liquefaction, and to permit the initial cool-down of large masses to about 80 K. Several arrangements for the pre-cooling process are possible depending on the desired application. Each arrangement has its proper advantages and drawbacks. The aim of this paper is to review the possible process solutions for liquid nitrogen pre-cooling and their particularities.

  8. Tritium release and retention properties of highly neutron-irradiated beryllium pebbles from HIDOBE-01 experiment

    Chakin, V., E-mail: vladimir.chakin@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rolli, R.; Moeslang, A.; Klimenkov, M.; Kolb, M.; Vladimirov, P.; Kurinskiy, P.; Schneider, H.-C. [Karlsruhe Institute of Technology, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 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)

    2013-11-15

    The current helium cooled pebble bed (HCPB) tritium breeding blanket concept for fusion reactors includes a bed of 1 mm diameter beryllium pebbles to act as a neutron multiplier. Beryllium pebbles, fabricated by the rotating electrode method, were neutron irradiated in the HFR in Petten within the HIDOBE-01 experiment. This study presents tritium release and retention properties and data on microstructure evolution of beryllium pebbles irradiated at 630, 740, 873, 948 K up to a damage dose of 18 dpa, corresponding to a helium accumulation of about 3000 appm. The measured cumulative released activity from the beryllium pebbles irradiated at 948 K was found to be significantly lower than the calculated value. After irradiation at 873 and 948 K scanning electron microscopy (SEM) and transmission electron microscopy (TEM) analyses revealed large pores or bubbles in the bulk and oxide films with a thickness of up to 8 μm at the surface of the beryllium pebbles. The radiation-enhanced diffusion of tritium and the formation of open porosity networks accelerate the tritium release from the beryllium pebbles during the high-flux neutron irradiation.

  9. Thermo-mechanical characterization of ceramic pebbles for breeding blanket

    Lo Frano, Rosa, E-mail: rosa.lofrano@ing.unipi.it; Aquaro, Donato; Scaletti, Luca

    2016-11-01

    Highlights: • Experimental activities to characterize the Li{sub 4}SiO{sub 4}. • Compression tests of pebbles. • Experimental evaluation of thermal conductivity of pebbles bed at different temperatures. • Experimental test with/without compression load. - Abstract: An open issue for fusion power reactor is to design a suitable breeding blanket capable to produce the necessary quantity of the tritium and to transfer the energy of the nuclear fusion reaction to the coolant. The envisaged solution called Helium-Cooled Pebble Bed (HCPB) breeding blanket foresees the use of lithium orthosilicate (Li{sub 4}SiO{sub 4}) or lithium metatitanate (Li{sub 2}TiO{sub 3}) pebble beds. The thermal mechanical properties of the candidate pebble bed materials are presently extensively investigated because they are critical for the feasibility and performances of the numerous conceptual designs which use a solid breeder. This study is aimed at the investigation of mechanical properties of the lithium orthosilicate and at the characterization of the main chemical, physical and thermo-mechanical properties taking into account the production technology. In doing that at the Department of Civil and Industrial Engineering (DICI) of the University of Pisa adequate experiments were carried out. The obtained results may contribute to characterize the material of the pebbles and to optimize the design of the envisaged fusion breeding blankets.

  10. Pressurized helium II-cooled magnet test facility

    Warren, R.P.; Lambertson, G.R.; Gilbert, W.S.; Meuser, R.B.; Caspi, S.; Schafer, R.V.

    1980-06-01

    A facility for testing superconducting magnets in a pressurized bath of helium II has been constructed and operated. The cryostat accepts magnets up to 0.32 m diameter and 1.32 m length with current to 3000 A. In initial tests, the volume of helium II surrounding the superconducting magnet was 90 liters. Minimum temperature reached was 1.7 K at which point the pumping system was throttled to maintain steady temperature. Helium II reservoir temperatures were easily controlled as long as the temperature upstream of the JT valve remained above T lambda; at lower temperatures control became difficult. Positive control of the temperature difference between the liquid and cold sink by means of an internal heat source appears necessary to avoid this problem. The epoxy-sealed vessel closures, with which we have had considerable experience with normal helium vacuum, also worked well in the helium II/vacuum environment

  11. Effect of wall thickness and helium cooling channels on duct magnetohydrodynamic flows

    He, Qingyun; Feng, Jingchao; Chen, Hongli

    2016-01-01

    Highlights: • MHD flows in ducts of different wall thickness compared with wall uniform. • Study of velocity, pressure distribution in ducts MHD flows with single pass of helium cooling channels. • Comparison of three types of dual helium cooling channels and acquisition of an option for minimum pressure drop. • A single short duct MHD flow in blanket without FCI has been simulated for pressure gradient analysis. - Abstract: The concept of dual coolant liquid metal (LM) blanket has been proposed in different countries to demonstrate the technical feasibility of DEMO reactor. In the system, helium gas and PbLi eutectic, separated by structure grid, are used to cool main structure materials and to be self-cooled, respectively. The non-uniform wall thickness of structure materials gives rise to wall non-homogeneous conductance ratio. It will lead to electric current distribution changes, resulting in significant changes in the velocity distribution and pressure drop of magnetohydrodynamic (MHD) flows. In order to investigate the effect of helium channels on MHD flows, different methods of numerical simulations cases are carried out including the cases of different wall thicknesses, single pass of helium cooling channels, and three types of dual helium cooling channels. The results showed that helium tubes are able to affect the velocity distribution in the boundary layer by forming wave sharp which transfers from Hartmann boundary layer to the core area. In addition, the potential profile and pressure drop in the cases have been compared to these in the case of walls without cooling channel, and the pressure gradient of a simplified single short duct MHD flow in blanket shows small waver along the central axis in the helium channel position.

  12. Effect of wall thickness and helium cooling channels on duct magnetohydrodynamic flows

    He, Qingyun; Feng, Jingchao; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn

    2016-02-15

    Highlights: • MHD flows in ducts of different wall thickness compared with wall uniform. • Study of velocity, pressure distribution in ducts MHD flows with single pass of helium cooling channels. • Comparison of three types of dual helium cooling channels and acquisition of an option for minimum pressure drop. • A single short duct MHD flow in blanket without FCI has been simulated for pressure gradient analysis. - Abstract: The concept of dual coolant liquid metal (LM) blanket has been proposed in different countries to demonstrate the technical feasibility of DEMO reactor. In the system, helium gas and PbLi eutectic, separated by structure grid, are used to cool main structure materials and to be self-cooled, respectively. The non-uniform wall thickness of structure materials gives rise to wall non-homogeneous conductance ratio. It will lead to electric current distribution changes, resulting in significant changes in the velocity distribution and pressure drop of magnetohydrodynamic (MHD) flows. In order to investigate the effect of helium channels on MHD flows, different methods of numerical simulations cases are carried out including the cases of different wall thicknesses, single pass of helium cooling channels, and three types of dual helium cooling channels. The results showed that helium tubes are able to affect the velocity distribution in the boundary layer by forming wave sharp which transfers from Hartmann boundary layer to the core area. In addition, the potential profile and pressure drop in the cases have been compared to these in the case of walls without cooling channel, and the pressure gradient of a simplified single short duct MHD flow in blanket shows small waver along the central axis in the helium channel position.

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

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

    2015-01-01

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

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

    Zheng Yanhua; Shi Lei; Wang Yan

    2010-01-01

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

  15. Performance of cold compressors in a cooling system of an R and D superconducting coil cooled with subcooled helium

    Hamaguchi, S.; Imagawa, S.; Yanagi, N.; Takahata, K.; Maekawa, R.; Mito, T.

    2006-01-01

    The helical coils of large helical device (LHD) have been operated in saturated helium at 4.4 K and plasma experiments have been carried out at magnetic fields lower than 3 T for 8 years. Now, it is considered that the cooling system of helical coils will be improved to enhance magnetic fields in 2006. In the improvement, the helical coils will be cooled with subcooled helium and the operating temperature of helical coils will be lowered to achieve the designed field of 3 T and enhance cryogenic stabilities. Two cold compressors will be used in the cooling system of helical coils to generate subcooled helium. In the present study, the performance of cold compressors has been investigated, using a cooling system of R and D coil, to apply cold compressors to the cooling system of helical coils. Actual surge lines of cold compressors were observed and the stable operation area was obtained. Automatic operations were also performed within the area. In the automatic operations, the suitable pressure of a saturated helium bath, calculated from the rotation speed of the 1st cold compressor, was regulated by bypass valve. From these results, stable operations will be expected in the cooling system of helical coils

  16. Cooling of nuclear power stations with high temperature reactors and helium turbine cycles

    Foerster, S.; Hewing, G.

    1977-01-01

    On nuclear power stations with high temperature reactors and helium turbine cycles (HTR-single circuits) the residual heat from the energy conversion process in the primary and intermediate coolers is removed from cycled gas, helium. Water, which is circulated for safety reasons through a closed circuit, is used for cooling. The primary and intermediate coolers as well as other cooling equipment of the power plant are installed within the reactor building. The heat from the helium turbine cycle is removed to the environment most effectively by natural draught cooling towers. In this way a net plant efficiency of about 40% is attainable. The low quantities of residual heat thereby produced and the high (in comparison with power stations with steam turbine cycles) cooling agent pressure and cooling water reheat pressure in the circulating coolers enable an economically favourable design of the overall 'cold end' to be expected. In the so-called unit range it is possible to make do with one or two cooling towers. Known techniques and existing operating experience can be used for these dry cooling towers. After-heat removal reactor shutdown is effected by a separate, redundant cooling system with forced air dry coolers. The heat from the cooling process at such locations in the power station is removed to the environment either by a forced air dry cooling installation or by a wet cooling system. (orig.) [de

  17. On the hyperporous non-linear elasticity model for fusion-relevant pebble beds

    Di Maio, P.A.; Giammusso, R.; Vella, G.

    2010-01-01

    Packed pebble beds are particular granular systems composed of a large amount of small particles, arranged in irregular lattices and surrounded by a gas filling interstitial spaces. Due to their heterogeneous structure, pebble beds have non-linear and strongly coupled thermal and mechanical behaviours whose constitutive models seem limited, being not suitable for fusion-relevant design-oriented applications. Within the framework of the modelling activities promoted for the lithiated ceramics and beryllium pebble beds foreseen in the Helium-Cooled Pebble Bed breeding blanket concept of DEMO, at the Department of Nuclear Engineering of the University of Palermo (DIN) a thermo-mechanical constitutive model has been set-up assuming that pebble beds can be considered as continuous, homogeneous and isotropic media. The present paper deals with the DIN non-linear elasticity constitutive model, based on the assumption that during the reversible straining of a pebble bed its effective logarithmic bulk modulus depends on the equivalent pressure according to a modified power law and its effective Poisson modulus remains constant. In these hypotheses the functional dependence of the effective tangential and secant bed deformation moduli on either the equivalent pressure or the volumetric strain have been derived in a closed analytical form. A procedure has been, then, defined to assess the model parameters for a given pebble bed from its oedometric test results and it has been applied to both polydisperse lithium orthosilicate and single size beryllium pebble beds.

  18. Is cold better ? - exploring the feasibility of liquid-helium-cooled optics

    Assoufid, L.; Mills, D.; Macrander, A.; Tajiri, G.

    1999-01-01

    Both simulations and recent experiments conducted at the Advanced Photon Source showed that the performance of liquid-nitrogen-cooled single-silicon crystal monochromators can degrade in a very rapid nonlinear fashion as the power and for power density is increased. As a further step towards improving the performance of silicon optics, we propose cooling with liquid helium, which dramatically improves the thermal properties of silicon beyond that of liquid nitrogen and brings the performance of single silicon-crystal-based synchrotrons radiation optics up to the ultimate limit. The benefits of liquid helium cooling as well as some of the associated technical challenges will be discussed, and results of thermal and structural finite elements simulations comparing the performance of silicon monochromators cooled with liquid nitrogen and helium will be given

  19. Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

    Larson, V. R.; Gunn, S. V.; Lee, J. C.

    1975-01-01

    The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

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

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

    2014-01-01

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

  1. European DEMO BOT Solid Breeder Blanket: the concept based on the use of cooling plates and beds of beryllium and Li4SiO4 pebbles

    Dalle Donne, M.; Fischer, U.; Norajitra, P.; Reimann, G.; Reiser, H.

    1995-01-01

    The paper presents an important modification of the European DEMO BOT Solid Breeder Blanket. The new design uses cooling plates rather than tubes. This allows a considerable simplification of the blanket and the separation of the beryllium from the Li 4 SiO 4 pebbles. The neutronic, thermohydraulic and tritium performance of the new design is quite good and equivalent to that of the previous one. (orig.)

  2. Core homogenization method for pebble bed reactors

    Kulik, V.; Sanchez, R.

    2005-01-01

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

  3. Helium release rates and ODH calculations from RHIC magnet cooling line failure

    Liaw, C.J.; Than, Y.; Tuozzolo, J.

    2011-03-28

    A catastrophic failure of the magnet cooling lines, similar to the LHC superconducting bus failure incident, could discharge cold helium into the RHIC tunnel and cause an Oxygen Deficiency Hazard (ODH) problem. A SINDA/FLUINT{reg_sign} model, which simulated the 4.5K/4 atm helium flowing through the magnet cooling system distribution lines, then through a line break into the insulating vacuum volumes and discharging via the reliefs into the RHIC tunnel, had been developed. Arc flash energy deposition and heat load from the ambient temperature cryostat surfaces are included in the simulations. Three typical areas: the sextant arc, the Triplet/DX/D0 magnets, and the injection area, had been analyzed. Results, including helium discharge rates, helium inventory loss, and the resulting oxygen concentration in the RHIC tunnel area, are reported. Good agreement had been achieved when comparing the simulation results, a RHIC sector depressurization test measurement, and some simple analytical calculations.

  4. Unexpected mobility of OH+ and OD+ molecular ions in cooled helium gas

    Isawa, R; Yamazoe, J; Tanuma, H; Ohtsuki, K

    2012-01-01

    Mobilities of OH + and OD + ions in cooled helium gas have been measured at gas temperature of 4.3 K. Measured mobilities of both ions as a function of an effective temperature T eff show a minimum around 80 K, and they are approaching to the polarization limits at very low T eff . These findings will be related to the extremely strong anisotropy of the interaction potential between the molecular ion and helium atom.

  5. Numerical modeling and validation of helium jet impingement cooling of high heat flux divertor components

    Koncar, Bostjan; Simonovski, Igor; Norajitra, Prachai

    2009-01-01

    Numerical analyses of jet impingement cooling presented in this paper were performed as a part of helium-cooled divertor studies for post-ITER generation of fusion reactors. The cooling ability of divertor cooled by multiple helium jets was analysed. Thermal-hydraulic characteristics and temperature distributions in the solid structures were predicted for the reference geometry of one cooling finger. To assess numerical errors, different meshes (hexagonal, tetra, tetra-prism) and discretisation schemes were used. The temperatures in the solid structures decrease with finer mesh and higher order discretisation and converge towards finite values. Numerical simulations were validated against high heat flux experiments, performed at Efremov Institute, St. Petersburg. The predicted design parameters show reasonable agreement with measured data. The calculated maximum thimble temperature was below the tile-thimble brazing temperature, indicating good heat removal capability of reference divertor design. (author)

  6. Manufacturing Technology of Ceramic Pebbles for Breeding Blanket

    Rosa Lo Frano

    2018-05-01

    Full Text Available An open issue for the fusion power reactor is the choice of breeding blanket material. The possible use of Helium-Cooled Pebble Breeder ceramic material in the form of pebble beds is of great interest worldwide as demonstrated by the numerous studies and research on this subject. Lithium orthosilicate (Li4SiO4 is a promising breeding material investigated in this present study because the neutron capture of Li-6 allows the production of tritium, 6Li (n, t 4He. Furthermore, lithium orthosilicate has the advantages of low activation characteristics, low thermal expansion coefficient, high thermal conductivity, high density and stability. Even if they are far from the industrial standard, a variety of industrial processes have been proposed for making orthosilicate pebbles with diameters of 0.1–1 mm. However, some manufacturing problems have been observed, such as in the chemical stability (agglomeration phenomena. The aim of this study is to provide a new methodology for the production of pebbles based on the drip casting method, which was jointly developed by the DICI-University of Pisa and Industrie Bitossi. Using this new (and alternative manufacturing technology, in the field of fusion reactors, appropriately sized ceramic pebbles could be produced for use as tritium breeders.

  7. Thermal and flow design of helium-cooled reactors

    Melese, G.; Katz, R.

    1984-01-01

    This book continues the American Nuclear Society's series of monographs on nuclear science and technology. Chapters of the book include information on the first-generation gas-cooled reactors; HTGR reactor developments; reactor core heat transfer; mechanical problems related to the primary coolant circuit; HTGR design bases; core thermal design; gas turbines; process heat HTGR reactors; GCFR reactor thermal hydraulics; and gas cooling of fusion reactors

  8. Studies on air ingress for pebble bed reactors

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

    2002-01-01

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

  9. Mechanical performance of irradiated beryllium pebbles

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

    Wei Renjie; Jian Hongbing

    1992-01-01

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

  11. High-power frequency-stabilized laser for laser cooling of metastable helium at 389 nm

    Koelemeij, J.C.J.; Hogervorst, W.; Vassen, W.

    2005-01-01

    A high-power, frequency-stabilized laser for cooling of metastable helium atoms using the 2 S13 →3 P23 transition at 389 nm has been developed. The 389 nm light is generated by frequency doubling of a titanium:sapphire laser in an external enhancement cavity containing a lithium-triborate nonlinear

  12. Characteristics of microstructure and tritium release properties of different kinds of beryllium pebbles for application in tritium breeding modules

    Kurinskiy, P., E-mail: petr.kurinskiy@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials – Applied Materials Physics (IAM-AWP), P.O. Box 3640, Karlsruhe 76021 (Germany); Vladimirov, P.; Moeslang, A. [Karlsruhe Institute of Technology, Institute for Applied Materials – Applied Materials Physics (IAM-AWP), P.O. Box 3640, Karlsruhe 76021 (Germany); Rolli, R. [Karlsruhe Institute of Technology, Institute for Applied Materials – Materials and Biomechanics (IAM-WBM), P.O. Box 3640, Karlsruhe 76021 (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, Barcelona 08019 (Spain)

    2014-10-15

    Highlights: • Tritium release properties and characteristics of microstructure of beryllium pebbles having different sizes of grains were studied. • Fine-grained beryllium pebbles showed the best ability to release tritium compared to pebbles from another charges. • Be pebbles with the grain sizes exceeding 100 μm contain a great number of small pores and inclusions presumably referring to the history of material fabrication. • The sizes of grains are one of a key characteristic of microstructure which influences the parameters of tritium release. - Abstract: Beryllium pebbles with diameters of 1 mm are considered to be perspective material for the use as neutron multiplier in tritium breeding modules of fusion reactors. Up to now, the design of helium-cooled breeding blanket in ITER project foresees the use of 1 mm beryllium pebbles fabricated by NGK Insulators Ltd., Japan. It is notable that beryllium pebbles from Russian Federation and USA are also available and the possibility of their large-scale fabrication is under study. Presented work is dedicated to a study of characteristics of microstructure and parameters of tritium release of beryllium pebbles produced by Bochvar Institute, Russian Federation, and Materion Corporation, USA.

  13. Porous structure analysis of large-scale randomly packed pebble bed in high temperature gas-cooled reactor

    Ren, Cheng; Yang, Xingtuan; Liu, Zhiyong; Sun, Yanfei; Jiang, Shengyao [Tsinghua Univ., Beijing (China). Key Laboratory of Advanced Reactor Engineering and Safety; Li, Congxin [Ministry of Environmental Protection of the People' s Republic of China, Beijing (China). Nuclear and Radiation Safety Center

    2015-02-15

    A three-dimensional pebble bed corresponding to the randomly packed bed in the heat transfer test facility built for the High Temperature Reactor Pebble bed Modules (HTR-PM) in Shandong Shidaowan is simulated via discrete element method. Based on the simulation, we make a detailed analysis on the packing structure of the pebble bed from several aspects, such as transverse section image, longitudinal section image, radial and axial porosity distributions, two-dimensional porosity distribution and coordination number distribution. The calculation results show that radial distribution of porosity is uniform in the center and oscillates near the wall; axial distribution of porosity oscillates near the bottom and linearly varies along height due to effect of gravity; the average coordination number is about seven and equals to the maximum coordination number frequency. The fully established three-dimensional packing structure analysis of the pebble bed in this work is of fundamental significance to understand the flow and heat transfer characteristics throughout the pebble-bed type structure.

  14. Thermal optimization of the helium-cooled power leads for the SSC

    Demko, J.A.; Schiesser, W.E.; Carcagno, R.; McAshan, M.; McConeghy, R.

    1992-01-01

    The optimum thermal design of the power leads for the Superconducting Super Collider (SSC) will minimize the amount of Carnot work (which is a combination of refrigeration and liquefaction work) required. This optimization can be accomplished by the judicious selection of lead length and diameter. Even though an optimum set of dimensions is found, the final design must satisfy other physical constraints such as maximum allowable heat leak and helium vapor mass flow rate. A set of corresponding lengths and diameters has been determined that meets these requirements for the helium vapor-cooled, spiral-fin power lead design of the SSC. Early efforts by McFee and Mallon investigated optimizing power leads for cryogenic applications with no convection cooling. Later designs utilized the boiled-off helium vapor to cool the lead. One notable design for currents up to several thousand amps is presented by Efferson based on a series of recommendations discussed by Deiness. Buyanov presents many theoretical models and design formulae but does not demonstrate an approach to thermally optimizing the design of a vapor-cooled lead. In this study, a detailed numerical thermal model of a power lead design for the SSC has been developed. It was adapted from the dynamic model developed by Schiesser. This model was used to determine the optimum dimensions that minimize the Carnot refrigeration and liquefaction work due to the leads. Since the SSC leads will be cooled by supercritical helium, the flow of vapor is regulated by a control valve. These leads include a superconducting portion at the cold end. All of the material properties in the model are functions of temperature, and for the helium are functions of pressure and temperature. No pressure drop calculations were performed as part of this analysis. The diameter that minimizes the Carnot work was determined for four different lengths at a design current of 6600 amps

  15. Thermal optimization of the helium-cooled power leads for the SSC

    Demko, J.A.; Schiesser, W.E.; Carcagno, R.; McAshan, M.; McConeghy, R.

    1992-03-01

    The optimum thermal design of the power leads for the Superconducting Super Collider (SSC) will minimize the amount of Carnot work (which is a combination of refrigeration and liquefaction work) required. This optimization can be accomplished by the judicious selection of lead length and diameter. Even though an optimum set of dimensions is found, the final design must satisfy other physical constraints such as maximum allowable heat leak and helium vapor mass flow rate. A set of corresponding lengths and diameters has been determined that meets these requirements for the helium vapor-cooled, spiral-fin power lead design of the SSC. Early efforts by McFee and Mallon investigated optimizing power leads for cryogenic applications with no convection cooling. Later designs utilized the boiled-off helium vapor to cool the lead. One notable design for currents up to several thousand amps is presented by Efferson based on a series of recommendations discussed by Deiness. Buyanov presents many theoretical models and design formulate but does not demonstrate an approach to thermally optimizing the design of a vapor-cooled lead. A method for optimizing superconducting magnet current leads is described by Maehata et al. The approach assumes that the helium boil-off caused by heat conduction along with power lead into the low-temperature helium is used to cool the lead. The optimum solution is found when the heat flow at the cold end is minimized.. In this study, a detailed numerical thermal model of a power lead design for the SSC has been developed. It was adapted from the dynamic model developed by Schiesser. This model was used to determine the optimum dimensions that minimize the Carnot refrigeration and liquefaction work due to the leads

  16. Cooling of superconducting electric generators by liquid helium

    Nakayama, W.; Ogata, H.

    1987-01-01

    Superconducting generators have a great potential in future electric supply systems in increasing the efficiency of generators and in enhancing the stability of power network systems. Recognition of possible advantages over gas-cooled and water-cooled generators has led research institutes and manufacturers in several countries to wage substantial research and development efforts. The authors show the electric power capacities of the test generators already built, under construction, or in the planning stage. Since earlier attempts, steady improvements in the design of generators have been made, and experience of generator operation has been accumulated

  17. Development of a neutronics code based on analytic function expansion nodal method for pebble-type High Temperature Gas-cooled Reactor design

    Cho, Nam Zin; Lee, Joo Hee; Lee, Jae Jun; Yu, Hui; Lee, Gil Soo [Korea Advanced Institute of Science and Tehcnology, Daejeon (Korea, Republic of)

    2006-03-15

    There is growing interest in developing Pebble Bed Reactors(PBRs) as a candidate of Very High Temperature gas-cooled Reactors(VHTRs). Until now, most existing methods of nuclear design analysis for this type of reactors are base on old finite-difference solvers or on statistical methods. And other existing nodal cannot be adapted for this kind of reactors because of transverse integration problem. In this project, we developed the TOPS code in three dimensional cylindrical geometry based on Analytic Function Expansion Nodal (AFEN) method developed at KAIST. The TOPS code showed better results in computing time than FDM and MCNP. Also TOPS showed very accurate results in reactor analysis.

  18. Development of a neutronics code based on analytic function expansion nodal method for pebble-type High Temperature Gas-cooled Reactor design

    Cho, Nam Zin; Lee, Joo Hee; Lee, Jae Jun; Yu, Hui; Lee, Gil Soo

    2006-03-01

    There is growing interest in developing Pebble Bed Reactors(PBRs) as a candidate of Very High Temperature gas-cooled Reactors(VHTRs). Until now, most existing methods of nuclear design analysis for this type of reactors are base on old finite-difference solvers or on statistical methods. And other existing nodal cannot be adapted for this kind of reactors because of transverse integration problem. In this project, we developed the TOPS code in three dimensional cylindrical geometry based on Analytic Function Expansion Nodal (AFEN) method developed at KAIST. The TOPS code showed better results in computing time than FDM and MCNP. Also TOPS showed very accurate results in reactor analysis

  19. The strategic study of pebble model high temperature gas-cooled reactor plant with power generation feature and industrial application prospect

    Zhao Mu; Ma Bo; Dong Yujie

    2010-01-01

    On the basis of the technical feature of pebble model high temperature gas-cooled reactor (HTR-PM) plant, its developmental advantage and future are deeply investigated from inherent safety and economics. It is explored about the business opportunity and future financing mode of HTR-PM plant. Industrial distribution and potential user are studied. It is resulted that the technical potential can be developed fully using Gas turbine power generation technology. It has wide market and great significance to build more group modules at home and developing countries. (authors)

  20. Thermal Performance of a Dual-Channel, Helium-Cooled, Tungsten Heat Exchanger

    Youchison, Dennis L.; North, Mart T.

    2000-01-01

    Helium-cooled, refractory heat exchangers are now under consideration for first wall and divertor applications. These refractory devices take advantage of high temperature operation with large delta-Ts to effectively handle high heat fluxes. The high temperature helium can then be used in a gas turbine for high-efficiency power conversion. Over the last five years, heat removal with helium was shown to increase dramatically by using porous metal to provide a very large effective surface area for heat transfer in a small volume. Last year, the thermal performance of a bare-copper, dual-channel, helium-cooled, porous metal divertor mock-up was evaluated on the 30 kW Electron Beam Test System at Sandia National Laboratories. The module survived a maximum absorbed heat flux of 34.6 MW/m 2 and reached a maximum surface temperature of 593 C for uniform power loading of 3 kW absorbed on a 2-cm 2 area. An impressive 10 kW of power was absorbed on an area of 24 cm 2 . Recently, a similar dual-module, helium-cooled heat exchanger made almost entirely of tungsten was designed and fabricated by Thermacore, Inc. and tested at Sandia. A complete flow test of each channel was performed to determine the actual pressure drop characteristics. Each channel was equipped with delta-P transducers and platinum RTDs for independent calorimetry. One mass flow meter monitored the total flow to the heat exchanger, while a second monitored flow in only one of the channels. The thermal response of each tungsten module was obtained for heat fluxes in excess of 5 MW/m 2 using 50 C helium at 4 MPa. Fatigue cycles were also performed to assess the fracture toughness of the tungsten modules. A description of the module design and new results on flow instabilities are also presented

  1. Particle exhaust of helium plasmas with actively cooled outboard pump limiter on Tore Supra

    Uckan, T.; Mioduszewski, P.K.; Loarer, T.; Chatelier, M.; Guilhem, D.; Lutz, T.; Nygren, R.E.; Mahdavi, M.A.

    1995-08-01

    The superconducting tokamak Tore Supra was designed for long-pulse (30-s) high input power operation. Here observations on the particle-handling characteristics of the actively cooled modular outboard pump limiter (OPL) are presented for helium discharges. The important experimental result was that a modest pumping speed (1 m 3 /s) of the OPL turbomolecular pump (TMP) provided background helium exhaust. This result came about due to a well-conditioned vessel wall with helium discharges that caused no wall outgasing. The particle accountability in these helium discharges was excellent, and the well-conditioned wall did not play a significant role in the particle balance. The helium density control, 25% density drop with OPL exhaust efficiency of ∼1%, was possible with TMP although this may not be the case with reactive gases such as deuterium. The observed quadratic increase of the OPL neutral pressure with helium density was consistent with an improvement of the particle control with increasing plasma density

  2. Conceptual design of two helium cooled fusion blankets (ceramic and liquid breeder) for INTOR

    Dalle Donne, M.; Dorner, S.; Taczanowski, S.

    1983-08-01

    Neutronic and heat transfer calculations have been performed for two helium cooled blankets for the INTOR design. The neutronic calculations show that the local tritium breeding ratios, both for the ceramic blanket (Li 2 SiO 3 ) and for the liquid blanket (Li 17 Pb 83 ) solutions, are 1.34 for natural tritium and about 1.45 using 30% Li 6 enrichment. The heat transfer calculations show that it is possible to cool the divertor section of the torus (heat flux = 1.7 MW/m 2 ) with helium with an inlet pressure of 52 bar and an inlet temperature of 40 0 C. The temperature of the back face of the divertor can be kept at 130 0 C. With helium with the same inlet conditions it is possible to cool the first wall as well (heat flux = 0.136 MW/m 2 ) and keep the back-face of this wall at a temperature of 120 0 C. For the ceramic blanket we use helium with 52 bar inlet pressure and 400 0 C inlet temperature to ensure sufficiently high temperatures in the breeder material. The maximum temperature in the pressure tubes containing the blanket is 450 0 C, while the maximum breeder particle temperature is 476 0 C. (orig./RW) [de

  3. Safety and environmental impact of the BOT helium cooled solid breeder blanket for DEMO. SEAL subtask 6.2, final report

    Kleefeldt, K.; Dammel, F.; Gabel, K.

    1996-03-01

    The European Union has been engaged since 1989 in a programme to develop tritium breeding blankets for application in a fusion power reactor. There are four concepts under development, namely two of the solid breeder type and two of the liquid breeder type. At the Forschungszentrum Karlsruhe one blanket concept of each line has been pursued so far with the so-called breeder outside tube (BOT) type representing the solid breeder line. In the BOT concept, Li 4 SiO 4 is used as ceramic breeding material in the form of pebble beds in combination with beryllium pebbles serving as neutron multiplier. Breeder and multiplier materials are arranged in radial-toroidal layers, separated by cooling plates. The coolant is high pressure helium which is circulated in series, at first through the first wall structure and subsequently through the cooling plates. The safety and environmental impact of the BOT blanket concept has been assessed as part of the blanket concept selection exercise, a European concerted action aiming at selecting the two most promising concepts for further development. The topics investigated are: (a) Blanket materials and toxic materials inventory, (b) energy sources for mobilisation, (c) fault tolerance, (d) tritium and activation product release, and (e) waste generation. No insurmountable safety problems have been identified for the BOT concept. The results of the assessment are described in this report. The information collected is also intended to serve as input to the EU 'Safety and Environmental Assessment of Fusion long-term Programme' (SEAL). The unresolved issues pertaining to the BOT blanket which need further investigations in future programmes are outlined herein. (orig.) [de

  4. Test results from a helium gas-cooled porous metal heat exchanger

    North, M.T.; Rosenfeld, J.H.; Youchison, D.L.

    1996-01-01

    A helium-cooled porous metal heat exchanger was built and tested, which successfully absorbed heat fluxes exceeding all previously tested gas-cooled designs. Helium-cooled plasma-facing components are being evaluated for fusion applications. Helium is a favorable coolant for fusion devices because it is not a plasma contaminant, it is not easily activated, and it is easily removed from the device in the event of a leak. The main drawback of gas coolants is their relatively poor thermal transport properties. This limitation can be removed through use of a highly efficient heat exchanger design. A low flow resistance porous metal heat exchanger design was developed, based on the requirements for the Faraday shield for the International Thermonuclear Experimental Reactor (ITER) device. High heat flux tests were conducted on two representative test articles at the Plasma Materials Test Facility (PMTF) at Sandia National Laboratories. Absorbed heat fluxes as high as 40 MW/m 2 were successfully removed during these tests without failure of the devices. Commercial applications for electronics cooling and other high heat flux applications are being identified

  5. A MEASUREMENT OF THE ADIABATIC COOLING INDEX FOR INTERSTELLAR HELIUM PICKUP IONS IN THE INNER HELIOSPHERE

    Saul, Lukas; Wurz, Peter; Kallenbach, Reinald

    2009-01-01

    Interstellar neutral gas enters the inner heliosphere where it is ionized and becomes the pickup ion population of the solar wind. It is often assumed that this population will subsequently cool adiabatically, like an expanding ideal gas due, to the divergent flow of the solar wind. Here, we report the first independent measure of the effective adiabatic cooling index in the inner heliosphere from SOHO CELIAS measurements of singly charged helium taken during times of perpendicular interplanetary magnetic field. We use a simple adiabatic transport model of interstellar pickup helium ions, valid for the upwind region of the inner heliosphere. The time averaged velocity spectrum of helium pickup ions measured by CELIAS/CTOF is fit to this model with a single free parameter which indicates an effective cooling rate with a power-law index of γ = 1.35 ± 0.2. While this average is consistent with the 'ideal-gas' assumption of γ = 1.5, the analysis indicates that such an assumption will not apply in general, and that due to observational constraints further measurements are necessary to constrain the cooling process. Implications are discussed for understanding the transport processes in the inner heliosphere and improving this measurement technique.

  6. The Cold Mass Support System and the Helium Cooling System for the MICE Focusing Solenoid

    Yang, Stephanie Q.; Green, Michael A.; Lau, Wing W.; Senanayake, Rohan S.; Witte, Holger

    2006-01-01

    The heart of the absorber focus coil (AFC) module for the muon ionization cooling experiment (MICE) is the two-coil superconducting solenoid that surrounds the muon absorber. The superconducting magnet focuses the muons that are cooled using ionization cooling, in order to improve the efficiency of cooling. The coils of the magnet may either be run in the solenoid mode (both coils operate at the same polarity) or the gradient (the coils operate at opposite polarity). The AFC magnet cold mass support system is designed to carry a longitudinal force up to 700 kN. The AFC module will be cooled using three pulse tube coolers that produce 1.5 W of cooling at 4.2 K. One of the coolers will be used to cool the liquid (hydrogen or helium) absorber used for ionization cooling. The other two coolers will cool the superconducting solenoid. This report will describe the MICE AFC magnet. The cold mass supports will be discussed. The reasons for using a pulsed tube cooler to cool this superconducting magnet will also be discussed

  7. Preliminary study on helium turbomachine for high temperature gas-cooled reactor

    Chen Yihua; Wang Jie; Zhang Zuoyi

    2003-01-01

    In the high temperature gas-cooled reactor (HTGR), gas turbine cycle is a new concept in the field of nuclear power. It combines two technologies of HTGR and gas turbine cycle, which represent the state-of-the-art technologies of nuclear power and fossil fuel generation respectively. This approach is expected to improve safety and economy of nuclear power plant significantly. So it is a potential scheme with competitiveness. The heat-recuperated cycle is the main stream of gas turbine cycle. In this cycle, the work medium is helium, which is very different from the air, so that the design features of the helium turbomachine and combustion gas turbomachine are different. The paper shows the basic design consideration for the heat-recuperated cycle as well as helium turbomachine and highlights its main design features compared with combustion gas turbomachine

  8. Study of Transient Heat Transport Mechanisms in Superfluid Helium Cooled Rutherford-Cables

    AUTHOR|(CDS)2100615

    The Large Hadron Collider leverages superconducting magnets to focus the particle beam or keep it in its circular track. These superconducting magnets are composed of NbTi-cables with a special insulation that allows superfluid helium to enter and cool the superconducting cable. Loss mechanisms, e.g. continuous random loss of particles escaping the collimation system heating up the magnets. Hence, a local temperature increase can occur and lead to a quench of the magnets when the superconductor warms up above the critical temperature. A detailed knowledge about the temperature increases in the superconducting cable (Rutherford type) ensures a secure operation of the LHC. A sample of the Rutherford cable has been instrumented with temperature sensors. Experiments with this sample have been performed within this study to investigate the cooling performance of the helium in the cable due to heat deposition. The experiment uses a superconducting coil, placed in a cryostat, to couple with the magnetic field loss m...

  9. Supercritical helium cooled, cabled, superconducting hollow conductors for large high field magnets

    Hoenig, M.O.; Iwasa, Y.; Montgomery, D.B.; Bejan, A.

    1976-01-01

    Within the last two years a new concept of cabled superconducting hollow conductors has been developed which are able to recover from transient instabilities by virtue of on-going, single-phase helium cooling. It has been possible to correlate small scale experimental results with an iterative computer program. The latter has been recently upgraded to include axial as well as radial heat transfer and predict more closely the chances of recovery. Nearly 1 g/s of supercritical helium has been circulated in a closed loop using a high speed centrifugal fan and up to 10 g/s using a reciprocating single pulse bellows pump. The loop is now being adapted to a 3 m length of a tightly wound 5000 A cabled hollow conductor equipped with pulse coils designed to fit inside a water cooled Bitter magnet. The combination will allow for a steady background field of 7.5 t with a 2 t superimposed pulse. (author)

  10. Fabrication and characterization of lithium orthosilicate pebbles using LiOH as a new raw material

    Knitter, R.; Reimann, J.; Risthaus, P.; Boccaccini, L.V.; Piazza, G.

    2004-01-01

    For the European Helium Cooled Pebble Bed (HCPB) blanket slightly overstoichiometric lithium orthosilicate pebbles (Li 4 SiO 4 +SiO 2 ) have been chosen as one optional breeder material. This material is developed in collaboration between Research Centre Karlsruhe (FZK) and the Schott Glas, Mainz. The lithium orthosilicate (OSi) pebbles are fabricated by the melting and spraying method in a semi-industrial scale facility. In the past, the not enriched pebbles were produced from a mixture of Li 4 SiO 4 and SiO 2 powders, but due to the fact that enriched Li 4 SiO 4 is not available on the market, highly enriched carbonate powder was used that finally resulted in nonsatisfying pebble characteristics. Enriched LiOH powder is commercially available, therefore, a new production route was pursued based on the following, simplified reaction: 4 LiOH + SiO 2 → Li 4 SiO 4 + 2 H 2 O. The melting process of LiOH and SiO 2 is less difficult to control than the melting of Li 2 CO 3 in spite of the decomposition of water. The pebbles produced from LiOH and SiO 2 are similar to those produced from Li 4 SiO 4 and SiO 2 . They exhibit a distinctly dendritic structure and show only a small amount of pores and cracks. In addition to the main constituent Li 4 SiO 4 , the high temperature phase Li 6 Si 2 O 7 was detected due to the quenching process and the excess of SiO 2 . This minor constituent, however, decomposes to Li 4 SiO 4 and Li 2 SiO 3 during annealing. In compressive crush load tests of single pebbles a crush load of about 9.5 N was measured for pebbles after drying at 300degC. The chemical analysis revealed a further advantage of the use of LiOH in the melting process. As LiOH is available in high-purity quality, the pebbles contain impurities to a lower degree than pebbles produced from Li 4 SiO 4 or Li 2 CO 3 . In order to obtain characteristic pebble bed data, first Uniaxial Compression Tests (UCTs) were performed at temperatures between ambient and at 850deg

  11. Design and safety analysis of the helium cooled solid breeder blanket for CFETR

    Wang, Shuai; Zhou, Guangming; Lv, Zhongliang; Jin, Cheng; Chen, Hongli [University of Science and Technology of China, Anhui (China). School of Nuclear Science and Technology

    2016-05-15

    This paper reports the design and safety analysis results of the helium cooled solid breeder blanket of the Chinese Fusion Engineering Test Reactor (CFETR). Materials selection and basic structure of the blanket have been presented. Performance analysis including neutronics analysis and thermo-mechanical analysis has shown good results. And the safety analysis of the blanket under Loss Of Coolant Accident (LOCA) conditions has been described. Results showed the current design can deal well with the selected accident scenarios.

  12. Researchers solve big mysteries of pebble bed reactor

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

    2014-03-15

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

  13. TEM study of impurity segregations in beryllium pebbles

    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.

  14. Localization of the Hot Spot in the Gap of Pebble Bed of Very High Temperature Gas Cooled Reactor(VHTGR)

    Lee, Sa Ya; Hong, Sung Je; Lee, Jae Young

    2010-01-01

    Pebble Bed Reactor(PBR) has been investigated intensively due to its benefits in management, but its complicated flow geometry requests reliable analytical methods. Hassan and Lee et al. have been made three dimensional computational methods. Hassan also measured local velocity fields with Particle Tracking Velocimetry(PTV), in small sized packed bed using liquid coolant, and Lee et al. measured flow field in the 2-dimensional wind tunnel with a hot wire system. In the present study, we develop the scaled up wind tunnel of pebble bed to use air as coolant in the same Reynolds number condition, as 21614, of the PBMR-250MWth. In order to measure the local surface temperature, the heating system and temperature measurement system were installed and heat transfer analogy was performed. The local surface temperature data shows that the predicted hot spots by Lee et al. at the top and bottom of the pebble by the velocity field measurement are reasonable, but the heat conduction is prior than contact effect at contact points

  15. Progress on pebble bed experimental activity for the HE-FUS3 mock-ups

    Dell'Orco, G.; Sansone, L.; Simoncini, M.; Zito, D.

    2002-01-01

    The EU Long Term for DEMO Programme foresees the qualification of the reference design of the helium cooled pebble bed (HCPB) - test blanket module (TBM) to be tested in ITER Reactor. In this frame, FZK and ENEA have launched many experimental activities for the evaluation of the interactions between the Tritium breeder and neutron multiplier pebble beds and the steel containment walls. Main aim of these activities is the measuring the pebble bed effective thermal conductivity, the wall heat transfer coefficient as well as their dependency from the mechanical constraints. The paper presents the progress of the testing activity and results of the tests on two mock-up, called Tazza and Helichetta, carried out on the HE-FUS3 facility at ENEA Brasimone. (orig.)

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

    Stroh, K.R.

    1979-03-01

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

  17. Electrical behaviour of ceramic breeder blankets in pebble form after γ-radiation

    E. Carella

    2015-07-01

    Full Text Available Lithium orthosilicate (Li4SiO4 ceramics in from of pebble bed is the European candidate for ITER testing HCPB (Helium Cooled Pebble Bed breeding modules. The breeder function and the shielding role of this material, represent the areas upon which attention is focused. Electrical measurements are proposed for monitoring the modification created by ionizing radiation and at the same time provide information on lithium movement in this ceramic structure. The electrical tests are performed on pebbles fabricated by Spray-dryer method before and after gamma-irradiation through a 60Co source to a fluence of 4.8 Gy/s till a total dose of 5 ∗ 105 Gy. The introduction of thermal annealing treatments during the electrical impedance spectroscopy (EIS measurements points out the recombination effect of the temperature on the γ-induced defects.

  18. Coupling a Supercritical Carbon Dioxide Brayton Cycle to a Helium-Cooled Reactor.

    Middleton, Bobby [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Pasch, James Jay [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kruizenga, Alan Michael [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Walker, Matthew [Sandia National Lab. (SNL-CA), Livermore, CA (United States)

    2016-01-01

    This report outlines the thermodynamics of a supercritical carbon dioxide (sCO2) recompression closed Brayton cycle (RCBC) coupled to a Helium-cooled nuclear reactor. The baseline reactor design for the study is the AREVA High Temperature Gas-Cooled Reactor (HTGR). Using the AREVA HTGR nominal operating parameters, an initial thermodynamic study was performed using Sandia's deterministic RCBC analysis program. Utilizing the output of the RCBC thermodynamic analysis, preliminary values of reactor power and of Helium flow rate through the reactor were calculated in Sandia's HelCO2 code. Some research regarding materials requirements was then conducted to determine aspects of corrosion related to both Helium and to sCO2 , as well as some mechanical considerations for pressures and temperatures that will be seen by the piping and other components. This analysis resulted in a list of materials-related research items that need to be conducted in the future. A short assessment of dry heat rejection advantages of sCO2> Brayton cycles was also included. This assessment lists some items that should be investigated in the future to better understand how sCO2 Brayton cycles and nuclear can maximally contribute to optimizing the water efficiency of carbon free power generation

  19. Design, fabrication, and testing of a helium-cooled module for the ITER divertor

    Baxi, C.B.; Smith, J.P.; Youchison, D.

    1994-08-01

    The International Thermonuclear Reactor (ITER) will have a single-null divertor with total power flow of 200 MW and a peak heat flux of about 5 MW/m 2 . The reference coolant for the divertor is water. However, helium is a viable alternative and offers advantages from safety considerations, such as excellent radiation stability and chemical inertness. In order to prove the feasibility of helium cooling at ITER relevant heat flux conditions, General Atomics designed, fabricated, and tested a helium-cooled divertor module. The module was made from dispersion strengthened copper, with a heat flux surface 25 mm wide and 80 mm long, designed for twice the ITER divertor heat flux. Different techniques were examined to enhance the heat transfer, which in turn reduced the flow and pumping power required to cool the module. It was concluded that an extended surface was the most practical solution. An optimization study was performed to find the best extended surface parameters. The optimum extended surface geometry consisted of fins: 10 mm high, 0.4 mm thick with a 1 mm pitch. It was estimated to require a pumping power of 150 W to remove 20 kW of power. This is more than an order of magnitude reduction in pumping power requirement, compared to smooth surface. The module was fabricated by electric discharge machining (EDM) process. The testing was carried out at SNLA during August 1993. The testing confirmed the design calculations. The peak heat flux during the test was 10 MW/m 2 applied over a surface area of 20 cm 2 . The pumping power calculated from flow rate and pressure drop measurement was about 160 W, which was less than 1% of the power removed. It is planned to test the module to higher temperature limits and higher heat fluxes during coming months. As a result of this effort we conclude that helium cooling of the ITER divertor is feasible without requiring a very large helium pressure or a large pumping power

  20. Optimization of mass-production conditions for tritium breeder pebbles based on slurry droplet wetting method

    Park, Yi-Hyun, E-mail: yhpark@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Min, Kyung-Mi; Ahn, Mu-Young; Cho, Seungyon; Lee, Young-Min [National Fusion Research Institute, Daejeon (Korea, Republic of); Park, Sang-Jin; Danish, Rehan; Lim, Chul-Hwan; Jo, Yong-Dae [IVT Co., Ltd., Daegu (Korea, Republic of)

    2016-11-01

    Highlights: • An automatic dispensing system was developed to improve uniformity and production rate of breeder pebbles. • The production rate of this system for Li{sub 2}TiO{sub 3} pebble was estimated at 50 kg/year. • The optimization of dispensing and sintering conditions for the mass-production of Li{sub 2}TiO{sub 3} pebble was conducted. • Integrity of Li{sub 2}TiO{sub 3} pebble was able to be ensured during mass-production process, especially during batch process. - Abstract: Lithium metatitanate (Li{sub 2}TiO{sub 3}) is being considered as tritium breeding material for solid-type breeding blanket, which are used in pebble-bed form. The total amount of Li{sub 2}TiO{sub 3} pebbles in Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) is approximately 80 kg. Furthermore, DEMO reactor requires a great deal of breeder pebbles. Therefore, the development of mass-production system for breeder pebbles is necessary. The slurry droplet wetting method was adopted in the mass-production process for Li{sub 2}TiO{sub 3} pebbles, which had been developed in Korea. In this method, an automatic slurry dispensing system is one of the key apparatuses because the uniformity of pebbles and production rate are able to be improved. The system was successfully manufactured, which was consisted of a dispensing unit for instillation of Li{sub 2}TiO{sub 3} slurry, a glycerin bath for hardening of droplets, and an automatic maintaining unit for constant distance between syringe needle and glycerin surface. The production rate of this system for Li{sub 2}TiO{sub 3} pebble was estimated at 50 kg/year. In this study, it was investigated that the effect of dispensing and sintering conditions on the mass-production of Li{sub 2}TiO{sub 3} pebbles.

  1. Light induced cooling of a heated solid immersed in liquid helium I

    Lezak, D.; Brodie, L.C.; Semura, J.S.

    1984-01-01

    This chapter investigates the marked enhancement in the transient heat transfer from the heater-thermometer to the liquid helium immediately following the application of a flash of visible light. This ''light effect'' is associated with increased bubble activity, and it is possible that the light induces a rapid nucleation of bubbles in the superheated liquid at or near the heater surface. A summary of the light effect is presented and some potential uses to which this effect could be applied are suggested. Quantification of the light effect and properties of the light effect are discussed. It is determined that the light effect is an additional cooling due to a light induced enhancement of boiling in superheated liquid helium I. The effect could be applied in practical cryogenic engineering and for the acquisition of fundamental knowledge of boiling heat transfer and nucleation in cryogenic liquids

  2. Options for a high heat flux enabled helium cooled first wall for DEMO

    Arbeiter, Frederik, E-mail: f.arbe@kit.edu; Chen, Yuming; Ghidersa, Bradut-Eugen; Klein, Christine; Neuberger, Heiko; Ruck, Sebastian; Schlindwein, Georg; Schwab, Florian; Weth, Axel von der

    2017-06-15

    Highlights: • Design challenges for helium cooled first wall reviewed and otimization approaches explored. • Application of enhanced heat transfer surfaces to the First Wall cooling channels. • Demonstrated a design point for 1 MW/m{sup 2} with temperatures <550 °C and acceptable stresses. • Feasibility of several manufacturing processes for ribbed surfaces is shown. - Abstract: Helium is considered as coolant in the plasma facing first wall of several blanket concepts for DEMO fusion reactors, due to the favorable properties of flexible temperature range, chemical inertness, no activation, comparatively low effort to remove tritium from the gas and no chemical corrosion. Existing blanket designs have shown the ability to use helium cooled first walls with heat flux densities of 0.5 MW/m{sup 2}. Average steady state heat loads coming from the plasma for current EU DEMO concepts are expected in the range of 0.3 MW/m{sup 2}. The definition of peak values is still ongoing and depends on the chosen first wall shape, magnetic configuration and assumptions on the fraction of radiated power and power fall off lengths in the scrape off layer of the plasma. Peak steady state values could reach and excess 1 MW/m{sup 2}. Higher short-term transient loads are expected. Design optimization approaches including heat transfer enhancement, local heat transfer tuning and shape optimization of the channel cross section are discussed. Design points to enable a helium cooled first wall capable to sustain heat flux densities of 1 MW/m{sup 2} at an average shell temperature lower than 500 °C are developed based on experimentally validated heat transfer coefficients of structured channel surfaces. The required pumping power is in the range of 3–5% of the collected thermal power. The FEM stress analyses show code-acceptable stress intensities. Several manufacturing methods enabling the application of the suggested heat transfer enhanced first wall channels are explored. An

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

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

    1987-01-01

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

  4. An evaluation of reactor cooling and coupled hydrogen production processes using the modular helium reactor

    Harvego, E.A.; Reza, S.M.M.; Richards, M.; Shenoy, A.

    2006-01-01

    The high-temperature characteristics of the modular helium reactor (MHR) make it a strong candidate for producing hydrogen using either thermochemical or high-temperature electrolysis (HTE) processes. Using heat from the MHR to drive a sulfur-iodine (SI) thermochemical hydrogen production process has been the subject of a U.S. Department of Energy sponsored Nuclear Engineering Research Initiative (NERI) project led by General Atomics, with participation from the Idaho National Laboratory (INL) and Texas A and M University. While the focus of much of the initial work was on the SI thermochemical production of hydrogen, recent activities included development of a preconceptual design for an integral HTE hydrogen production plant driven by the process heat and electricity produced by a 600 MW MHR. This paper describes ATHENA analyses performed to evaluate alternative primary system cooling configurations for the MHR to minimize peak reactor vessel and core temperatures while achieving core helium outlet temperatures in the range of 900-1000 deg. C that are needed for the efficient production of hydrogen using either the SI or HTE process. The cooling schemes investigated are intended to ensure peak fuel temperatures do not exceed specified limits under normal or transient upset conditions, and that reactor vessel temperatures do not exceed American Society of Mechanical Engineers (ASME) code limits for steady-state or transient conditions using standard light water reactor vessel materials. Preconceptual designs for SI and HTE hydrogen production plants driven by one or more 600 MW MHRs at helium outlet temperatures in the range of 900-1000 deg. C are described and compared. An initial SAPHIRE model to evaluate the reliability, maintainability, and availability of the SI hydrogen production plant is also described. Finally, a preliminary flowsheet for a conceptual design of an HTE hydrogen production plant coupled to a 600 MW modular helium reactor is presented and

  5. Conceptual design and analysis of the helium cooled solid breeder blanket for CFETR

    Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Li, Min; Lv, Zhongliang; Zhou, Guangming; Liu, Qianwen; Wang, Shuai; Wang, Xiaoliang; Zheng, Jie; Ye, Minyou

    2015-10-15

    Highlights: • A helium cooled solid blanket was proposed as a candidate blanket concept for CFETR. • Material selection, basic structure and gas flow scheme of the blanket were introduced. • A series of performance analyses for the blanket were summarized. - Abstract: To bridge the gap between ITER and DEMO and to realize the fusion energy in China, a fusion device Chinese Fusion Engineering Test Reactor (CFETR) was proposed and is being designed mainly to demonstrate 50–200 MW fusion power, 30–50% duty time factor, tritium self-sustained. Because of the high demand of tritium production and the realistic engineering consideration, the design of tritium breeding blanket for CFETR is a challenging work and getting special attention. As a blanket candidate, a helium cooled solid breeder blanket has been designed with the emphasis on conservative design and realistic blanket technology. This paper introduces the basic blanket scheme, including the material selection, structural design, cooling scheme and purge gas flow path. In addition, some results of neutronics, thermal-hydraulic and stress analysis are presented.

  6. Gas-cooled fast-breeder reactor. Helium Circulator Test Facility updated design cost estimate

    1979-04-01

    Costs which are included in the cost estimate are: Titles I, II, and III Architect-Engineering Services; Titles I, II, and III General Atomic Services; site clearing, grading, and excavation; bulk materials and labor of installation; mechanical and electrical equipment with installation; allowance for contractors' overhead, profit, and insurance; escalation on materials and labor; a contingency; and installation of GAC supplied equipment and materials. The total estimated cost of the facility in As Spent Dollars is $27,700,000. Also included is a cost comparison of the updated design and the previous conceptual design. There would be a considerable penalty for the direct-cooled system over the indirect-cooled system due to the excessive cost of the large diameter helium loop piping to an outdoor heat exchanger. The indirect cooled system which utilizes a helium/Dowtherm G heat exchanger and correspondingly smaller and lower pressure piping to its outdoor air cooler proved to be the more economical of the two systems

  7. Random geometry capability in RMC code for explicit analysis of polytype particle/pebble and applications to HTR-10 benchmark

    Liu, Shichang; Li, Zeguang; Wang, Kan; Cheng, Quan; She, Ding

    2018-01-01

    Highlights: •A new random geometry was developed in RMC for mixed and polytype particle/pebble. •This capability was applied to the full core calculations of HTR-10 benchmark. •Reactivity, temperature coefficient and control rod worth of HTR-10 were compared. •This method can explicitly model different packing fraction of different pebbles. •Monte Carlo code with this method can simulate polytype particle/pebble type reactor. -- Abstract: With the increasing demands of high fidelity neutronics analysis and the development of computer technology, Monte Carlo method is becoming more and more attractive in accurate simulation of pebble bed High Temperature gas-cooled Reactor (HTR), owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. For the double-heterogeneous geometry of pebble bed, traditional Monte Carlo codes can treat it by explicit geometry description. However, packing methods such as Random Sequential Addition (RSA) can only produce a sphere packing up to 38% volume packing fraction, while Discrete Element Method (DEM) is troublesome and also time consuming. Moreover, traditional Monte Carlo codes are difficult and inconvenient to simulate the mixed and polytype particles or pebbles. A new random geometry method was developed in Monte Carlo code RMC to simulate the particle transport in polytype particle/pebble in double heterogeneous geometry systems. This method was verified by some test cases, and applied to the full core calculations of HTR-10 benchmark. The reactivity, temperature coefficient and control rod worth of HTR-10 were compared for full core and initial core in helium and air atmosphere respectively, and the results agree well with the benchmark results and experimental results. This work would provide an efficient tool for the innovative design of pebble bed, prism HTRs and molten salt reactors with polytype particles or pebbles using Monte Carlo method.

  8. Modelling of thermal and mechanical behaviour of pebble beds

    Boccaccini, L.V.; Buehler, L.; Hermsmeyer, S.; Wolf, F.

    2001-01-01

    FZK (Forshungzentrum Karlsruhe) is developing a Helium Cooled Pebble Bed (HCPB) Blanket Concept for fusion power reactors based on the use of ceramic breeder materials and beryllium multiplier in the form of pebble beds. The design of such a blanket requires models and computer codes describing the thermal-mechanical behavior of pebble beds to evaluate the temperatures, stresses, deformations and mechanical interactions between pebble beds and the structure with required accuracy and reliability. The objective to describe the beginning of life condition for the HCPB blanket seems near to be reached. Mechanical models that describe the thermo-mechanical behavior of granular materials used in form of pebble beds are implemented in a commercial structure code. These models have been calibrated using the results of a large series of dedicated experiments. The modeling work is practically concluded for ceramic breeder; it will be carried on in the next year for beryllium to obtain the required correlations for creep and the thermal conductivity. The difficulties for application in large components (such as the HCPB blanket) are the limitations of the present commercial codes to manage such a set of constitutive equations under complex load conditions and large mesh number. The further objective is to model the thermal cycles during operation; the present correlations have to be adapted for the release phase. A complete description of the blanket behavior during irradiation is at the present out of our capability; this objective requires an extensive R and D program that at the present is only at the beginning. (Y.Tanaka)

  9. Computer simulation of multiple stability regions in an internally cooled superconducting conductor and of helium replenishment in a bath-cooled conductor

    Turner, L.R.; Shindler, J.

    1984-09-01

    For upcoming fusion experiments and future fusion reactors, superconducting magnetic have been chosen or considered which employ cooling by pool-boiling HeI, by HeII, and by internally flowing HeI. The choice of conductor and cooling method should be determined in part by the response of the magnet to sudden localized heat pulses of various magnitudes. The paper describes the successful computer simulation of multiple stability in internally cooled conductors, as observed experimentally, using the computer code SSICC. It also describes the modeling of helium replenishment in the cooling channels of a bath-cooled conductor, using the computer code TASS

  10. Reduction of circulation power for helium-cooled fusion reactor blanket using additive CO{sub 2} gas

    Lee, Yeon-Gun [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Department of Nuclear and Energy Engineering, Jeju National University, 102 Jejudaehakno, Jeju-si 690-756, Jeju (Korea, Republic of); Park, Il-Woong [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Lee, Dong Won [Nuclear Fusion Engineering Development Center, Korea Atomic Energy Research Institute, Daedeokdaero 989 beon-gil, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Kim, Eung-Soo, E-mail: kes7741@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of)

    2015-11-15

    Helium (He) cooling requires large circulation power to remove high heat from plasma side and nuclear heating by high energy neutron in fusion reactors due to its low density. Based on the recent findings that the heat transfer capability of the light gas can be enhanced by mixing another heavier gas, this study adds CO{sub 2} to a reference helium coolant and evaluates the cooling performance of the binary mixture for various compositions. To assess the cooling performance, computational fluid dynamic (CFD) analyses on the KO HCML (Korea Helium Cooled Molten Lithium) TBM are conducted. As a result, it is revealed that the binary mixing of helium, which has favorable thermophysical properties but the density, with a heavier noble gas or an unreactive gas significantly reduces the required circulation power by an order of magnitude with meeting the thermal design requirements. This is attributed to the fact that the density can be highly increased with small amount of a heavier gas while other gas properties are kept relatively comparable. The optimal CO{sub 2} mole fraction is estimated to be 0.4 and the circulation power, in this case, can be reduced to 13% of that of pure helium. This implies that the thermal efficiency of a He-cooled blanket system can be fairly enhanced by means of the proposed binary mixing.

  11. A robust helium-cooled shield/blanket design for ITER

    Wong, C.P.C.; Bourque, R.F.; Baxi, C.B.

    1993-11-01

    General Atomics Fusion and Reactor Groups have completed a helium-cooled, conceptual shield/blanket design for ITER. The configuration selected is a pressurized tubes design embedded in radially oriented plates. This plate can be made from ferritic steel or from V-alloy. Helium leakage to the plasma chamber is eliminated by conservative, redundant design and proper quality control and inspection programs. High helium pressure at 18 MPa is used to reduce pressure drop and enhance heat transfer. This high gas pressure is believed practical when confined in small diameter tubes. Ample industrial experience exists for safe high gas pressure operations. Inboard shield design is highlighted in this study since the allowable void fraction is more limited. Lithium is used as the thermal contacting medium and for tritium breeding, its safety concerns are minimized by a modular, low inventory design that requires no circulation of the liquid metal for the purpose of heat removal. This design is robust, conservative, reliable, and meets all design goals and requirements. It can also be built with present-day technology

  12. The Design of High Reliability Magnetic Bearing Systems for Helium Cooled Reactor Machinery

    Swann, M.; Davies, N.; Jayawant, R.; Leung, R.; Shultz, R.; Gao, R.; Guo, Z.

    2014-01-01

    The requirements for magnetic bearing equipped machinery used in high temperature, helium cooled, graphite moderated reactor applications present a set of design considerations that are unlike most other applications of magnetic bearing technology in large industrial rotating equipment, for example as used in the oil and gas or other power generation applications. In particular, the bearings are typically immersed directly in the process gas in order to take advantage of the design simplicity that comes about from the elimination of ancillary lubrication and cooling systems for bearings and seals. Such duty means that the bearings will usually see high temperatures and pressures in service and will also typically be subject to graphite particulate and attendant radioactive contamination over time. In addition, unlike most industrial applications, seismic loading events become of paramount importance for the magnetic bearings system, both for actuators and controls. The auxiliary bearing design requirements, in particular, become especially demanding when one considers that the whole mechanical structure of the magnetic bearing system is located inside an inaccessible pressure vessel that should be rarely, if ever, disassembled over the service life of the power plant. Lastly, many machinery designs for gas cooled nuclear power plants utilize vertical orientation. This circumstance presents its own unique requirements for the machinery dynamics and bearing loads. Based on the authors’ experience with machine design and supply on several helium cooled reactor projects including Ft. St. Vrain (US), GT-MHR (Russia), PBMR (South Africa), GTHTR (Japan), and most recently HTR-PM (China), this paper addresses many of the design considerations for such machinery and how the application of magnetic bearings directly affects machinery reliability and availability, operability, and maintainability. Remote inspection and diagnostics are a key focus of this paper. (author)

  13. Preliminary Analysis on Decay Heat Removal Capability of Helium Cooled Solid Breeder Test Blanket Module

    Ahn, Mu Young; Cho, Seung Yon; Kim, Duck Hoi; Lee, Eun Seok; Kim, Hyung Seok; Suh, Jae Seung; Yun, Sung Hwan; Cho, Nam Zin

    2007-01-01

    One of the main ITER goals is to test and validate design concepts of tritium breeding blankets relevant to DEMO or fusion power plants. Korea Helium-Cooled Solid Breeder (HCSB) Test Blanket Module (TBM) has been developed with overall objectives of achieving this goal. The TBM employs high pressure helium to cool down the First Wall (FW), Side Wall (SW) and Breeding Zone (BZ). Therefore, safety consideration is a part of the design process. Each ITER Party performing the TBM program is requested to reach a similar level of confidence in the TBM safety analysis. To meet ITER's request, Failure Mode and Effects Analysis (FMEA) studies have been performed on the TBM to identify the Postulated Initial Event (PIE). Although FMEA on the KO TBM has not been completed, in-vessel, in-box and ex-vessel Loss Of Coolant Accident (LOCA) are considered as enveloping cases of PIE in general. In this paper, accidental analyses for the three selected LOCA were performed to investigate the decay heat removal capability of the TBM. To simulate transient thermo-hydraulic behavior of the TBM for the selected scenarios, RELAP5/MOD3.2 code was used

  14. Updated conceptual design of helium cooling ceramic blanket for HCCB-DEMO

    Wang, Suhao [University of Science and Technology of China, Hefei, Anhui (China); Southwestern Institute of Physics, Chengdu, Sichuan (China); Cao, Qixiang; Wu, Xinghua; Wang, Xiaoyu; Zhang, Guoshu [Southwestern Institute of Physics, Chengdu, Sichuan (China); Feng, Kaiming, E-mail: fengkm@swip.ac.cn [Southwestern Institute of Physics, Chengdu, Sichuan (China)

    2016-11-15

    Highlights: • An updated design of Helium Cooled Ceramic breeder Blanket (HCCB) for HCCB-DEMO is proposed in this paper. • The Breeder Unit is transformed to TBM-like sub-modules, with double “banana” shape tritium breeder. Each sub-module is inserted in space formed by Stiffen Grids (SGs). • The performance analysis is performed based on the R&D development of material, fabrication technology and safety assessment in CN ITER TBM program. • Hot spots will be located at the FW bend side. - Abstract: The basic definition of the HCCB-DEMO plant and preliminary blanket designed by Southwestern Institution of Physics was proposed in 2009. The DEMO fusion power is 2550 MW and electric power is 800 MW. Based on development of R&D in breeding blanket, a conceptual design of helium cooled blanket with ceramic breeder in HCCB-DEMO was presented. The main design features of the HCCB-DEMO blanket were: (1) CLF-1 structure materials, Be multiplier and Li{sub 4}SiO{sub 4} breeder; (2) neutronic wall load is 2.3 MW/m{sup 2} and surface heat flux is 0.43 MW/m{sup 2} (2) TBR ≈ 1.15; (3) geometry of breeding units is ITER TBM-like segmentation; (4)Pressure of helium is 8 MPa and inlet/outlet temperature is 300/500 °C. On the basis of these design, some important analytical results are presented in aspects of (i) neutronic behavior of the blanket; (ii) design of 3D structure and thermal-hydraulic lay-out for breeding blanket module; (iii) structural-mechanical behavior of the blanket under pressurization. All of these assessments proved current stucture fulfill the design requirements.

  15. High-heat-flux testing of helium-cooled heat exchangers for fusion applications

    Youchison, D.L.; Izenson, M.G.; Baxi, C.B.; Rosenfeld, J.H.

    1996-01-01

    High-heat-flux experiments on three types of helium-cooled divertor mock-ups were performed on the 30-kW electron beam test system and its associated helium flow loop at Sandia National Laboratories. A dispersion-strengthened copper alloy (DSCu) was used in the manufacture of all the mock-ups. The first heat exchanger provides for enhanced heat transfer at relatively low flow rates and much reduced pumping requirements. The Creare sample was tested to a maximum absorbed heat flux of 5.8 MW/m 2 . The second used low pressure drops and high mass flow rates to achieve good heat removal. The GA specimen was tested to a maximum absorbed heat flux of 9 MW/m 2 while maintaining a surface temperature below 400 degree C. A second experiment resulted in a maximum absorbed heat flux of 34 MW/m 2 and surface temperatures near 533 degree C. The third specimen was a DSCu, axial flow, helium-cooled divertor mock-up filled with a porous metal wick which effectively increases the available heat transfer area. Low mass flow and high pressure drop operation at 4.0 MPa were characteristic of this divertor module. It survived a maximum absorbed heat flux of 16 MW/m 2 and reached a surface temperature of 740 degree C. Thermacore also manufactured a follow-on, dual channel porous metal-type heat exchanger, which survived a maximum absorbed heat flux of 14 MW/m 2 and reached a maximum surface temperature of 690 degree C. 11refs., 20 figs., 3 tabs

  16. Helium circulator design concepts for the modular high temperature gas-cooled reactor (MHTGR) plant

    McDonald, C.F.; Nichols, M.K.; Kaufman, J.S.

    1988-01-01

    Two helium circulators are featured in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) power plant - (1) the main circulator, which facilitates the transfer of reactor thermal energy to the steam generator, and (2) a small shutdown cooling circulator that enables rapid cooling of the reactor system to be realized. The 3170 kW(e) main circulator has an axial flow compressor, the impeller being very similar to the unit in the Fort St. Vrain (FSV) plant. The 164 kW(e) shutdown cooling circulator, the design of which is controlled by depressurized conditions, has a radial flow compressor. Both machines are vertically oriented, have submerged electric motor drives, and embody rotors that are supported on active magnetic bearings. As outlined in this paper, both machines have been conservatively designed based on established practice. The circulators have features and characteristics that have evolved from actual plant operating experience. With a major goal of high reliability, emphasis has been placed on design simplicity, and both machines are readily accessible for inspection, repair, and replacement, if necessary. In this paper, conceptual design aspects of both machines are discussed, together with the significant technology bases. As appropriate for a plant that will see service well into the 21st century, new and emerging technologies have been factored into the design. Examples of this are the inclusion of active magnetic bearings, and an automated circulator condition monitoring system. (author). 18 refs, 20 figs, 13 tabs

  17. Postirradiation examination of beryllium pebbles

    Gelles, D.S.

    1998-01-01

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

  18. Fabrication of Li{sub 2}TiO{sub 3} pebbles using PVA–boric acid reaction for solid breeding materials

    Park, Yi-Hyun, E-mail: yhpark@nfri.re.kr; Cho, Seungyon; Ahn, Mu-Young

    2014-12-15

    Highlights: • Li{sub 2}TiO{sub 3} 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 Li{sub 2}TiO{sub 3} pebble was exceptionally homogeneous. • Suitable process conditions for high-quality Li{sub 2}TiO{sub 3} pebble were summarized. - Abstract: Lithium metatitanate (Li{sub 2}TiO{sub 3}) 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, Li{sub 2}TiO{sub 3} 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 Li{sub 2}TiO{sub 3} green body were investigated. In addition, the basic characteristics of the sintered pebble were also evaluated. The shape of Li{sub 2}TiO{sub 3} green bodies was affected by slurry viscosity, PVA content and boric acid content. The grain size and average crush load of sintered Li{sub 2}TiO{sub 3} pebble were controlled by the sintering time. The boron was completely removed during the final sintering process.

  19. Pebble bed pebble motion: Simulation and Application

    Cogliati, Joshua J.

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

  20. Helium pressures in RHIC vacuum cryostats and relief valve requirements from magnet cooling line failure

    Liaw, C.J.; Than, Y.; Tuozzolo, J.

    2011-03-28

    A catastrophic failure of the RHIC magnet cooling lines, similar to the LHC superconducting bus failure incident, would pressurize the insulating vacuum in the magnet and transfer line cryostats. Insufficient relief valves on the cryostats could cause a structural failure. A SINDA/FLUINT{reg_sign} model, which simulated the 4.5K/4 atm helium flowing through the magnet cooling system distribution lines, then through a line break into the vacuum cryostat and discharging via the reliefs into the RHIC tunnel, had been developed to calculate the helium pressure inside the cryostat. Arc flash energy deposition and heat load from the ambient temperature cryostat surfaces were included in the simulations. Three typical areas: the sextant arc, the Triplet/DX/D0 magnets, and the injection area, had been analyzed. Existing relief valve sizes were reviewed to make sure that the maximum stresses, caused by the calculated maximum pressures inside the cryostats, did not exceed the allowable stresses, based on the ASME Code B31.3 and ANSYS results. The conclusions are as follows: (1) The S/F simulation results show that the highest internal pressure in the cryostats, due to the magnet line failure, is {approx}37 psig (255115 Pa); (2) Based on the simulation, the temperature on the cryostat chamber, INJ Q8-Q9, could drop to 228 K, which is lower than the material minimum design temperature allowed by the Code; (3) Based on the ASME Code and ANSYS results, the reliefs on all the cryostats inside the RHIC tunnel are adequate to protect the vacuum chambers when the magnet cooling lines fail; and (4) In addition to the pressure loading, the thermal deformations, due to the temperature decrease on the cryostat chambers, could also cause a high stress on the chamber, if not properly supported.

  1. Helium pressures in RHIC vacuum cryostats and relief valve requirements from magnet cooling line failure

    Liaw, C.J.; Than, Y.; Tuozzolo, J.

    2011-01-01

    A catastrophic failure of the RHIC magnet cooling lines, similar to the LHC superconducting bus failure incident, would pressurize the insulating vacuum in the magnet and transfer line cryostats. Insufficient relief valves on the cryostats could cause a structural failure. A SINDA/FLUINT(reg s ign) model, which simulated the 4.5K/4 atm helium flowing through the magnet cooling system distribution lines, then through a line break into the vacuum cryostat and discharging via the reliefs into the RHIC tunnel, had been developed to calculate the helium pressure inside the cryostat. Arc flash energy deposition and heat load from the ambient temperature cryostat surfaces were included in the simulations. Three typical areas: the sextant arc, the Triplet/DX/D0 magnets, and the injection area, had been analyzed. Existing relief valve sizes were reviewed to make sure that the maximum stresses, caused by the calculated maximum pressures inside the cryostats, did not exceed the allowable stresses, based on the ASME Code B31.3 and ANSYS results. The conclusions are as follows: (1) The S/F simulation results show that the highest internal pressure in the cryostats, due to the magnet line failure, is ∼37 psig (255115 Pa); (2) Based on the simulation, the temperature on the cryostat chamber, INJ Q8-Q9, could drop to 228 K, which is lower than the material minimum design temperature allowed by the Code; (3) Based on the ASME Code and ANSYS results, the reliefs on all the cryostats inside the RHIC tunnel are adequate to protect the vacuum chambers when the magnet cooling lines fail; and (4) In addition to the pressure loading, the thermal deformations, due to the temperature decrease on the cryostat chambers, could also cause a high stress on the chamber, if not properly supported.

  2. PBMR Project - Pebble Fuel Advantages

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

    2006-01-01

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

  3. RF DEMO ceramic helium cooled blanket, coolant and energy transformation systems

    Kovalenko, V.; Leshukov, A.; Poliksha, V.; Popov, A.; Strebkov, Yu.; Borisov, A.; Shatalov, G.; Demidov, V.; Kapyshev, V.

    2004-01-01

    RF DEMO-S reactor is a prototype of commercial fusion reactors for further generation. A blanket is the main element unit of the reactor design. The segment structure is the basis of the ceramic blanket. The segments mounting/dismounting operations are carried out through the vacuum vessel vertical port. The inboard/outboard blanket segment is the modules welded design, which are welded by back plate. The module contains the back plate, the first wall, lateral walls and breeding zone. The 9CrMoVNb steel is used as structural material. The module internal space formed by the first wall, lateral walls and back plate is used for breeding zone arrangement. The breeding zone design based upon the poloidal BIT (Breeder Inside Tube) concept. The beryllium is used as multiplier material and the lithium orthosilicate is used as breeder material. The helium at 0.1 MPa is used as purge gas. The cooling is provided by helium at 10 MPa. The coolant supply/return to the blanket modules are carrying out on the two independent circuits. The performed investigations of possible transformation schemes of DEMO-S blanket heat power into the electricity allowed to make a conclusion about the preferable using of traditional steam-turbine facility in the secondary circuit. (author)

  4. Study of Temperature Wave Propagation in Superfluid Helium Focusing on Radio-Frequency Cavity Cooling

    Koettig, T; Avellino, S; Junginger, T; Bremer, J

    2015-01-01

    Oscillating Superleak Transducers (OSTs) can be used to localize quenches of superconducting radio-frequency cavities. Local hot spots at the cavity surface initiate temperature waves in the surrounding superfluid helium that acts as cooling fluid at typical temperatures in the range of 1.6 K to 2 K. The temperature wave is characterised by the properties of superfluid helium such as the second sound velocity. For high heat load densities second sound velocities greater than the standard literature values are observed. This fast propagation has been verified in dedicated small scale experiments. Resistors were used to simulate the quench spots under controlled conditions. The three dimensional propagation of second sound is linked to OST signals. The aim of this study is to improve the understanding of the OST signal especially the incident angle dependency. The characterised OSTs are used as a tool for quench localisation on a real size cavity. Their sensitivity as well as the time resolution was proven to b...

  5. Activation analysis and waste management of China ITER helium cooled solid breeder test blanket module

    Han, J.R., E-mail: hanjingru@163.co [North China Electric Power University, School of Nuclear Science and Engineering, Zhu-Xin-Zhuang, De-Wai, Beijing 102206 (China); Chen, Y.X.; Han, R. [North China Electric Power University, School of Nuclear Science and Engineering, Zhu-Xin-Zhuang, De-Wai, Beijing 102206 (China); Feng, K.M. [Southwestern Institute of Physics, P.O.Box 432, Chengdu 610041 (China); Forrest, R.A. [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon (United Kingdom)

    2010-08-15

    Activation characteristics have been assessed for the ITER China helium cooled solid breeder (CH-HCSB) 3 x 6 test blanket module (TBM). Taking a representative irradiation scenario, the activation calculations were performed by FISPACT code. Neutron fluxes distributions in the TBM were provided by a preceding MCNP calculation. These fluxes were passed to FISPACT for the activation calculation. The main activation parameters of the HCSB-TBM were calculated and discussed, such as activity, afterheat and contact dose rate. Meanwhile, the dominant radioactivity nuclides and reaction channel pathways have been identified. According to the Safety and Environmental Assessment of Fusion Power (SEAFP) waste management strategy, the activated materials can be re-used following the remote handling recycling options. The results will provide useful indications for further optimization design and waste management of the TBM.

  6. Numerical benchmark for the deep-burn modular helium-cooled reactor (DB-MHR)

    Taiwo, T. A.; Kim, T. K.; Buiron, L.; Varaine, F.

    2006-01-01

    Numerical benchmark problems for the deep-burn concept based on the prismatic modular helium-cooled reactor design (a Very High Temperature Reactor (VHTR)) are specified for joint analysis by U.S. national laboratories and industry and the French CEA. The results obtained with deterministic and Monte Carlo codes have been inter-compared and used to confirm the underlying feature of the DB-MHR concept (high transuranics consumption). The results are also used to evaluate the impact of differences in code methodologies and nuclear data files on the code predictions for DB-MHR core physics parameters. The code packages of the participating organizations (ANL and CEA) are found to give very similar results. (authors)

  7. Thermal-hydraulic investigations on the CEA-ENEA DEMO relevant helium cooled poloidal blanket

    Dell'Orco, G.; Polazzi, G.; Vallette, F.; Proust, E.; Eid, M.

    1994-01-01

    The CEA-ENEA design of an Helium Cooled Solid Breeder Blanket (HCSBB) for the DEMO reactor, with a breeder in tube (BIT) poloidal arrangement, is based on the use of lithium ceramic pellets, the ENEA γ-LiAlO 2 or the CEA Li 2 ZrO 3 . Due to the geometry of the DEMO reactor plasma chamber, these breeder bundles are adapted to the Vacuum Vessel with a strong poloidal curvature. This curvature influences the thermal-hydraulic behaviour of the coolant flowing inside the bundle. The paper presents the CEA-ENEA first results of the experimental and theoretical programme, aiming at optimizing the breeder module thermal hydraulic design. (author) 6 refs.; 7 figs.; 1 tab

  8. Cryogenic thermometer calibration system using a helium cooling loop and a temperature controller [for LHC magnets

    Chanzy, E; Thermeau, J P; Bühler, S; Joly, C; Casas-Cubillos, J; Balle, C

    1998-01-01

    The IPN-Orsay and CERN are designing in close collaboration a fully automated cryogenic thermometer calibration facility which will calibrate in 3 years 10,000 cryogenic thermometers required for the Large Hadron Collider (LHC) operation. A reduced-scale model of the calibration facility has been developed, which enables the calibration of ten thermometers by comparison with two rhodium-iron standard thermometers in the 1.8 K to 300 K temperature range under vacuum conditions. The particular design, based on a helium cooling loop and an electrical temperature controller, gives good dynamic performances. This paper describes the experimental set-up and the data acquisition system. Results of experimental runs are also presented along with the estimated global accuracy for the calibration. (3 refs).

  9. Considerations on techniques for improving tritium confinement in helium-cooled ceramic breeder blankets

    Fuetterer, M.A.; Raepsaet, X.; Proust, E.; Leger, D.

    1994-01-01

    Tritium control issues such as the development of permeation barriers and the choice of the coolant and purge-gas chemistry are of crucial importance for solid breeder blankets. In order to quantify these problems for the helium-cooled ceramic breeder-inside-tube (BIT) blanket concept, the tritium leakage into the coolant was evaluated and the consequent tritium losses into the steam circuit were determined. The results indicate that under certain specified conditions the total tritium release from the coolant can be limited to approximately 10 Ci/d, but only on the assumption that experimental data for tritium permeation barriers can be attained under realistic operating conditions. An experimental study on the impact of the gas chemistry on tritium losses is proposed. (author) 8 refs.; 2 figs

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

    Zarins, Arturs, E-mail: arturs.zarins@lu.lv [University of Latvia, Institute of Chemical Physics, Jelgavas Street 1, LV-1004, Riga (Latvia); University of Latvia, Faculty of Chemistry, Jelgavas Street 1, LV-1004, Riga (Latvia); Valtenbergs, Oskars [University of Latvia, Institute of Chemical Physics, Jelgavas Street 1, LV-1004, Riga (Latvia); University of Latvia, Faculty of Chemistry, Jelgavas Street 1, LV-1004, Riga (Latvia); Kizane, Gunta; Supe, Arnis [University of Latvia, Institute of Chemical Physics, Jelgavas Street 1, LV-1004, Riga (Latvia); Knitter, Regina; Kolb, Matthias H.H.; Leys, Oliver [Karlsruhe Institute of Technology, Institute for Applied Materials (IAM-KWT), 76021, Karlsruhe (Germany); Baumane, Larisa [University of Latvia, Institute of Chemical Physics, Jelgavas Street 1, LV-1004, Riga (Latvia); Latvian Institute of Organic Synthesis, Aizkraukles Street 21, LV-1006, Riga (Latvia); Conka, Davis [University of Latvia, Institute of Chemical Physics, Jelgavas Street 1, LV-1004, Riga (Latvia); University of Latvia, Faculty of Chemistry, Jelgavas Street 1, LV-1004, Riga (Latvia)

    2016-03-15

    Lithium orthosilicate (Li{sub 4}SiO{sub 4}) pebbles with 2.5 wt.% excess of silicon dioxide (SiO{sub 2}) 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 Li{sub 4}SiO{sub 4} 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 Li{sub 4}SiO{sub 4} pebbles and to add titanium dioxide (TiO{sub 2}), to obtain lithium metatitanate (Li{sub 2}TiO{sub 3}) 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 Li{sub 4}SiO{sub 4} pebbles with different contents of TiO{sub 2} for the first time, in order to estimate and compare radiation stability. The reference and the modified Li{sub 4}SiO{sub 4} 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 Li{sub 4}SiO{sub 4} pebbles, several paramagnetic RD and RP are formed and accumulated, like, E' centres (SiO{sub 3}{sup 3−}/TiO{sub 3}{sup 3−}), HC{sub 2} centres (SiO{sub 4}{sup 3−}/TiO{sub 3}{sup −}) etc. On the basis of the obtained results, it is concluded that the modified Li{sub 4}SiO{sub 4} pebbles with TiO{sub 2} additions have comparable radiation stability with the reference pebbles. - Highlights: • Formation of RD and RP in modified Li{sub 4}SiO{sub 4} pebbles with additions of TiO{sub 2} is analysed for the first time. • Due to additions of TiO{sub 2}, concentration of paramagnetic RD slightly increased in modified Li{sub 4}SiO{sub 4} pebbles. • Modified Li{sub 4}SiO{sub 4} pebbles have good radiation stability compared to

  11. A helium-cooled blanket design of the low aspect ratio reactor

    Wong, C.P.; Baxi, C.B.; Reis, E.E.; Cerbone, R.; Cheng, E.T.

    1998-03-01

    An aggressive low aspect ratio scoping fusion reactor design indicated that a 2 GW(e) reactor can have a major radius as small as 2.9 m resulting in a device with competitive cost of electricity at 49 mill/kWh. One of the technology requirements of this design is a high performance high power density first wall and blanket system. A 15 MPa helium-cooled, V-alloy and stagnant LiPb breeder first wall and blanket design was utilized. Due to the low solubility of tritium in LiPb, there is the concern of tritium migration and the formation of V-hydride. To address these issues, a lithium breeder system with high solubility of tritium has been evaluated. Due to the reduction of blanket energy multiplication to 1.2, to maintain a plant Q of > 4, the major radius of the reactor has to be increased to 3.05 m. The inlet helium coolant temperature is raised to 436 C in order to meet the minimum V-alloy temperature limit everywhere in the first wall and blanket system. To enhance the first wall heat transfer, a swirl tape coolant channel design is used. The corresponding increase in friction factor is also taken into consideration. To reduce the coolant system pressure drop, the helium pressure is increased from 15 to 18 MPa. Thermal structural analysis is performed for a simple tube design. With an inside tube diameter of 1 cm and a wall thickness of 1.5 mm, the lithium breeder can remove an average heat flux and neutron wall loading of 2 and 8 MW/m(2), respectively. This reference design can meet all the temperature and material structural design limits, as well as the coolant velocity limits. Maintaining an outlet coolant temperature of 650 C, one can expect a gross closed cycle gas turbine thermal efficiency of 45%. This study further supports the use of helium coolant for high power density reactor design. When used with the low aspect ratio reactor concept a competitive fusion reactor can be projected at 51.9 mill/kWh

  12. MIT pebble bed reactor project

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

    2007-03-15

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

  13. MIT pebble bed reactor project

    Kadak, Andrew C.

    2007-01-01

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

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

    Bolier, E; Ferrero, C. [Forschungszentrum Karlsruhe, Zimer 203, Gebaeude 451, Abteilung HVT-TL (Germany); Moslang, A. [Forschungszentrum Karlsruhe GmbH, FZK, Karlsruhe (Germany); Pieritz, R.A. [CNRS, Lab. de Glaciologie et Geophysique de l' Environnement, 38 - Saint Martin d' Heres (France)

    2007-07-01

    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 T{sub 2} 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 10{sup 25} nxm{sup -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 {mu}m voxel resolution, respectively, followed by morphological and topological post

  15. Development of gas cooled reactors and experimental setup of high temperature helium loop for in-pile operation

    Miletić, Marija, E-mail: marija_miletic@live.com [Czech Technical University in Prague, Prague (Czech Republic); Fukač, Rostislav, E-mail: fuk@cvrez.cz [Research Centre Rez Ltd., Rez (Czech Republic); Pioro, Igor, E-mail: Igor.Pioro@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada); Dragunov, Alexey, E-mail: Alexey.Dragunov@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada)

    2014-09-15

    Highlights: • Gas as a coolant in Gen-IV reactors, history and development. • Main physical parameters comparison of gas coolants: carbon dioxide, helium, hydrogen with water. • Forced convection in turbulent pipe flow. • Gas cooled fast reactor concept comparisons to very high temperature reactor concept. • High temperature helium loop: concept, development, mechanism, design and constraints. - Abstract: Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuel supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow's needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a very high temperature reactor (VHTR). The VHTR concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behavior within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Center Rez Ltd. One of the topics analyzed in this article are also physical characteristic and benefits of gas

  16. Experimental study of the critical density of heat flux in open channels cooled with helium - II

    Pron'ko, V.G.; Gorokhov, V.V.; Saverin, V.N.

    1981-01-01

    Experimental values of the critical density of a heat flux qsub(cr) in uniformly heated open channels cooled with helium-2 are reported for the first time. The experimental test bench and experimental element are described. Experimental data are obtained in cylindrical channels of 12Kh18N1OT steel with inner diameter d=0.8, 1.8; 2.8 mm and ratio l/d=20.8, 44, 85. The channel orientation has varied from vertical to horizontal position, the immersion depth - from 100, to 600 mm. It has been found that the heat transfer crisis propagation over the whole length of the channel with He-2 occurs practically instantaneously. The qsub(cr) value depends essentially on the bath liquid temperature, angle of inclivnation and relative length (l/d) of the channel with qsub(cr) approximately (l/d)sup(-1.5) being independent of the depth of channel immersion. The obtained values of critical density of a heat flux in channels are papproximately by an order less than those found for a great bulk of He-2. The results presented may be used for designing various types of devices cooled with He-2 and development of heat exchange theory in it [ru

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

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

    2015-10-15

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

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

    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.

  19. CARR-CNS with crescent-shape moderator cell and sub-cooling helium jacket around cell

    Yu, Qingfeng; Feng, Quanke; Kawai, Takeshi; Cheng, Liang; Shen, Feng; Yuan, Luzheng

    2005-01-01

    The new type of the moderator cell was developed for the Cold Neutron Source (CNS) of the China Advanced Research Reactor (CARR) which is now constructing at the China Institute of Atomic Energy in Beijing. A crescent-shape moderator cell covered by the sub-cooling helium jacket is adopted. A crescent-shape would help to increase the volume of the moderator cell for corresponding it to the 4 cold neutron guide tubes, even if liquid hydrogen not liquid deuterium were used as a cold moderator. The sub-cooling helium jacket covering the moderator cell removes the nuclear heating of the outer shell wall of the cell. It contributes to reduce the void fraction of liquid hydrogen in the inner shell. Such a type of a moderator cell is suitable for the CNS with higher nuclear heating. The cold helium gas flows down firstly into the sub-cooling helium jacket and then flows up to the condenser. Therefore, the theory of the self-regulation for the thermo-siphon type of the CNS is also applicable

  20. Strength analysis of CARR-CNS with crescent-shape moderator cell and helium sub-cooling jacket covering cell

    Yu Qingfeng; Feng Quanke; Kawai Takeshi; Shen Feng; Yuan Luzheng; Cheng Liang

    2005-01-01

    The new type of the moderator cell was developed for the cold neutron source (CNS) of the China Advanced Research Reactor (CARR) which is now being constructed at the China Institute of Atomic Energy in Beijing. A crescent-shape moderator cell covered by the helium sub-cooling jacket is adopted. The structure of the moderator cell is optimized by the stress FEM analysis. A crescent-shape would help to increase the volume of the moderator cell for fitting it to the four cold neutron guide tubes, even if liquid hydrogen, not liquid deuterium, was used as a cold moderator. The helium sub-cooling jacket covering the moderator cell removes the nuclear heating of the outer shell wall of the cell. It contributes to reduce the void fraction of liquid hydrogen in the outer shell of the moderator cell. Such a type of a moderator cell is suitable for the CNS with higher nuclear heating. The cold helium gas flows down first into the helium sub-cooling jacket and then flows up to the condenser. The theory of the self-regulation suitable to the thermo-siphon type of the CNS is also applicable and validated

  1. CARR-CNS with crescent-shape moderator cell and sub-cooling helium jacket surrounding cell

    Yu, Qingfeng; Feng, Quanke; Kawai, Takeshi; Shen, Feng; Yuan, Luzheng

    2005-01-01

    The new type of the moderator cell was developed for the Cold Neutron Source (CNS) of the China Advanced Research Reactor (CARR) which is now constructing at the China Institute of Atomic Energy in Beijing. A crescent-shape moderator cell covered by the sub-cooling helium jacket is adopted. A crescent-shape would help to increase the volume of the moderator cell for corresponding it to the 4 cold neutron guide tubes, even if liquid hydrogen not liquid deuterium were used as a cold moderator. The sub-cooling helium jacket covering the moderator cell removes the nuclear heating of the outer shell wall of the cell. It contributes to reduce the void fraction of liquid hydrogen in the inner shell. Such a type of a moderator cell is suitable for the CNS with higher nuclear heating. The cold helium gas flows down firstly into the sub-cooling helium jacket and then flows up to the condenser. Therefore, the theory of the self-regulation for the thermo-siphon type of the CNS is also applicable

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

    Boer, Brian

    2007-01-01

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

  3. Status on DEMO Helium Cooled Lithium Lead breeding blanket thermo-mechanical analyses

    Aubert, J., E-mail: julien.aubert@cea.fr [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Aiello, G.; Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Kiss, B. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Budapest (Hungary); Morin, A. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France)

    2016-11-01

    Highlights: • CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. The DEMO HCLL breeding blanket design capitalizes on the experience acquired on the HCLL Test Blanket Module designed for ITER. Design improvements are being implemented to adapt the design to DEMO specifications and performance objectives. • Thermal and mechanical analyses have been carried out in order to justify the design of the HCLL breeding blanket showing promising results for tie rods modules’ attachments system and relatively good behavior of the box in case of LOCA when comparing to RCC-MRx criteria. • CFD thermal analyses on generic breeding unit have enabled the consolidation of the results obtained with previous FEM design analyses. - Abstract: The EUROfusion Consortium develops a design of a fusion power demonstrator (DEMO) in the framework of the European “Horizon 2020” innovation and research program. One of the key components in the fusion reactor is the breeding blanket surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. The Helium Cooled Lithium Lead (HCLL) blanket is one of the concepts which is investigated for DEMO. It is made of a Eurofer structure and uses the eutectic liquid lithium–lead as tritium breeder and neutron multiplier, and helium gas as coolant. Within the EUROfusion organization, CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. This paper presents the status of the thermal and mechanical analyses carried out on the HCLL breeding blanket in order to justify the design. CFD thermal analyses on generic breeding unit including stiffening plates and cooling plates have been performed with ANSYS in order to consolidate results obtained with previous FEM design analyses. Moreover in order to expand the justification of the HCLL Breeding blanket design, the most loaded area of

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

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

    2014-01-01

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

  5. Cryodeposition of nitrogen gas on a surface cooled by helium II

    Dhuley, R. C.; Bosque, E. S.; Van Sciver, S. W.

    2014-01-01

    Catastrophic loss of beam tube vacuum in a superconducting particle accelerator can be simulated by sudden venting of a long high vacuum channel cooled on its outer surface by He II. The rapid rush of atmospheric air in such an event shows an interesting propagation effect, which is much slower than the shock wave that occurs with vacuum loss at ambient conditions. This is due to flash frosting/deposition of air on the cold walls of the channel. Hence to characterize the propagation as well as the associated heat transfer, it is first necessary to understand the deposition process. Here we attempt to model the growth of nitrogen frost layer on a cold plate in order to estimate its thickness with time. The deposition process can be divided into two regimes- free molecular and continuum. It is shown that in free molecular regime, the frost growth can be modeled reasonably well using cryopump theory and general heat transfer relations. The continuum regime is more complex to model, given the higher rate of gas incident on cryosurface causing a large heat load on helium bath and changing cryosurface temperature. Results from the continuum regime are discussed in the context of recent experiments performed in our laboratory

  6. Cryodeposition of nitrogen gas on a surface cooled by helium II

    Dhuley, R. C.; Bosque, E. S.; Van Sciver, S. W. [Cryogenics Group, National High Magnetic Field Laboratory, Tallahassee, FL 32310 USA and Mechanical Engineering Department, FAMU-FSU College of Engineering, Tallahassee, FL 32310 (United States)

    2014-01-29

    Catastrophic loss of beam tube vacuum in a superconducting particle accelerator can be simulated by sudden venting of a long high vacuum channel cooled on its outer surface by He II. The rapid rush of atmospheric air in such an event shows an interesting propagation effect, which is much slower than the shock wave that occurs with vacuum loss at ambient conditions. This is due to flash frosting/deposition of air on the cold walls of the channel. Hence to characterize the propagation as well as the associated heat transfer, it is first necessary to understand the deposition process. Here we attempt to model the growth of nitrogen frost layer on a cold plate in order to estimate its thickness with time. The deposition process can be divided into two regimes- free molecular and continuum. It is shown that in free molecular regime, the frost growth can be modeled reasonably well using cryopump theory and general heat transfer relations. The continuum regime is more complex to model, given the higher rate of gas incident on cryosurface causing a large heat load on helium bath and changing cryosurface temperature. Results from the continuum regime are discussed in the context of recent experiments performed in our laboratory.

  7. Post-examination of helium-cooled tungsten components exposed to DEMO specific cyclic thermal loads

    Ritz, G.; Hirai, T.; Linke, J.; Norajitra, P.; Giniyatulin, R.; Singheiser, L.

    2009-01-01

    A concept of helium-cooled tungsten finger module was developed for the European DEMO divertor. The concept was realized and tested under DEMO specific cyclic thermal loads up to 10 MW/m 2 . The modules were examined carefully before and after loading by metallography and microstructural analyses. While before loading mainly discrete and shallow cracks were found on the tungsten surface due to the manufacturing process, dense crack networks were observed at the loaded surfaces due to the thermal stress. In addition, cracks occurred in the structural, heat sink part and propagated along the grains orientation of the deformed tungsten material. Facilitated by cracking, the molten brazing metal between the tungsten plasma facing material and the W-La 2 O 3 heat sink, that could not withstand the operational temperatures, infiltrated the tungsten components and, due to capillary forces, even reached the plasma facing surface through the cracks. The formed cavity in the brazed layer reduced the heat conduction and the modules were further damaged due to overheating during the applied heat loads. Based on this detailed characterization and possible improvements of the design and of the manufacturing routes are discussed.

  8. Brookhaven program to develop a helium-cooled power transmission system

    Forsyth, E.B.

    1975-01-01

    The particular system under design consists of flexible cables installed in a cryogenic enclosure at room temperature and cooled to the range 6 to 9 0 K by supercritical helium, contraction of the cable is accommodated by proper choice of helix angles of the components of the cable. The superconductor is Nb 3 Sn and at the present time the dielectric insulation is still the subject of intensive development. Two good choices appear to be forms of polyethylene and polycarbonate. Sample cables incorporating various dielectrics have been manufactured commercially in lengths of 1500 ft and tested in laboratory cryostats in shorter sections of about 70 ft. A test facility is under construction to evaluate cables and cryogenic components for this type of service, the first refrigerator uses a 350 H.P. screw compressor and three turbo-expander stages. It is hoped to achieve reliability of a very high order. The first three-phase tests will be conducted at 69 kV, although it appears that 230 to 345 kV is the most likely voltage range for future applications. (auth)

  9. Conceptual design of the blanket mechanical attachment for the helium-cooled lithium-lead reactor

    Barrera, G.; Branas, B.; Lucas, J.; Doncel, J.; Medrano, M.; Garcia, A.; Giancarli, L.; Ibarra, A.; Li Puma, A.; Maisonnier, D.; Sardain, P.

    2008-01-01

    The conceptual design of a new type of fusion reactor based on the helium-cooled lithium-lead (HCLL) blanket has been performed within the European Power Plant Conceptual Studies. As part of this activity, a new attachment system suitable for the HCLL blanket modules had to be developed. This attachment is composed of two parts. The first one is the connection between module and the first part of a shield, called high temperature shield, which operates at a temperature around 500 deg. C, close to that of the blanket module. This connection must be made at the lateral walls, in order to avoid openings through the first wall and breeding zone thus avoiding complex design and fabrication issues of the module. The second connection is the one between the high temperature shield and a second shield called low temperature shield, which has a temperature during reactor operation around 150 deg. C. The design of this connection is complex because it must allow the large differential thermal expansion (up to 30 mm) between the two components. Design proposals for both connections are presented, together with the results of finite element mechanical analyses which demonstrate the feasibility to support the blanket and shield modules during normal and accidental operation conditions

  10. Electromagnetic analysis on Korean Helium Cooled Ceramic Reflector (HCCR) TBM during plasma major disruption

    Lee, Youngmin; Ku, Duck Young; Ahn, Mu-Young; Cho, Seungyon; Park, Yi-Hyun; Lee, Dong Won

    2015-01-01

    Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) will be installed at the #18 equatorial port of the Vaccum Vessel in order to test the feasibility of the breeding blanket performance for forthcoming fusion power plant in the ITER TBM Program. Since ITER tokamak contains Vaccum Vessel and set of electromagnetic coils, the TBM as well as other components is greatly influenced by magnetic field generated by these coils. By the electromagnetic (EM) fast transient events such as major disruption (MD), vertical displacement event (VDE) or magnet fast discharge (MFD) occurred in tokamak system, the eddy current can be induced eventually in the conducting components. As a result, the magnetic field and induced eddy current produce extremely huge EM load (force and moment) on the TBM. Therefore, EM load calculation is one of the most important analyses for optimized design of TBM. In this study, a 20-degree sector model for tokamak system including central solenoid (CS) coil, poloidal field (PF) coil, toroidal field (TF) coil, vaccum vessel, shield blankets and TBM set (TBM, TBM key, TBM shield, TBM frame) is prepared for analysis by ANSYS-EMAG tool. Concerning the installation location of the TBM, a major disruption scenario is particularly applied for fast transient analysis. The final goal of this study is to evaluate the EM load on HCCR TBM during plasma major disruption.

  11. Electromagnetic analysis on Korean Helium Cooled Ceramic Reflector (HCCR) TBM during plasma major disruption

    Lee, Youngmin, E-mail: ymlee@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Ku, Duck Young; Ahn, Mu-Young; Cho, Seungyon; Park, Yi-Hyun [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) will be installed at the #18 equatorial port of the Vaccum Vessel in order to test the feasibility of the breeding blanket performance for forthcoming fusion power plant in the ITER TBM Program. Since ITER tokamak contains Vaccum Vessel and set of electromagnetic coils, the TBM as well as other components is greatly influenced by magnetic field generated by these coils. By the electromagnetic (EM) fast transient events such as major disruption (MD), vertical displacement event (VDE) or magnet fast discharge (MFD) occurred in tokamak system, the eddy current can be induced eventually in the conducting components. As a result, the magnetic field and induced eddy current produce extremely huge EM load (force and moment) on the TBM. Therefore, EM load calculation is one of the most important analyses for optimized design of TBM. In this study, a 20-degree sector model for tokamak system including central solenoid (CS) coil, poloidal field (PF) coil, toroidal field (TF) coil, vaccum vessel, shield blankets and TBM set (TBM, TBM key, TBM shield, TBM frame) is prepared for analysis by ANSYS-EMAG tool. Concerning the installation location of the TBM, a major disruption scenario is particularly applied for fast transient analysis. The final goal of this study is to evaluate the EM load on HCCR TBM during plasma major disruption.

  12. Pumped helium system for cooling positron and electron traps to 1.2 K

    Wrubel, J; Kolthammer, W S; Larochelle, P; McConnell, R; Richerme, P; Grzonka, D; Oelert, W; Sefzick, T; Zielinski, M; Borbely, J S; George, M C; Hessels, E A; Storry, C H; Weel, M; Mullers, A; Walz, J; Speck, A

    2011-01-01

    Extremely precise tests of fundamental particle symmetries should be possible via laser spectroscopy of trapped antihydrogen ((H) over bar) atoms. (H) over bar atoms that can be trapped must have an energy in temperature units that is below 0.5 K-the energy depth of the deepest magnetic traps that can currently be constructed with high currents and superconducting technology. The number of atoms in a Boltzmann distribution with energies lower than this trap depth depends sharply upon the temperature of the thermal distribution. For example, ten times more atoms with energies low enough to be trapped are in a thermal distribution at a temperature of 1.2 K than for a temperature of 4.2 K. To date, (H) over bar atoms have only been produced within traps whose electrode temperature is 4.2 K or higher. A lower temperature apparatus is desirable if usable numbers of atoms that can be trapped are to eventually be produced. This report is about the pumped helium apparatus that cooled the trap electrodes of an (H) ove...

  13. Non-linear Model Predictive Control for cooling strings of superconducting magnets using superfluid helium

    AUTHOR|(SzGeCERN)673023; Blanco Viñuela, Enrique

    In each of eight arcs of the 27 km circumference Large Hadron Collider (LHC), 2.5 km long strings of super-conducting magnets are cooled with superfluid Helium II at 1.9 K. The temperature stabilisation is a challenging control problem due to complex non-linear dynamics of the magnets temperature and presence of multiple operational constraints. Strong nonlinearities and variable dead-times of the dynamics originate at strongly heat-flux dependent effective heat conductivity of superfluid that varies three orders of magnitude over the range of possible operational conditions. In order to improve the temperature stabilisation, a proof of concept on-line economic output-feedback Non-linear Model Predictive Controller (NMPC) is presented in this thesis. The controller is based on a novel complex first-principles distributed parameters numerical model of the temperature dynamics over a 214 m long sub-sector of the LHC that is characterized by very low computational cost of simulation needed in real-time optimizat...

  14. Activation analysis of Chinese ITER helium cooled solid breeder test blanket module

    Han Jingru; Chen Yixue; Ma Xubo; Wang Shouhai; Forrest, R.A.

    2009-01-01

    Based on the Chinese ITER helium cooled solid breeder(CH-HCSB) test blanket module (TBM) of the 3 x 6 sub-modules options, the activation characteristics of the TBM were calculated. Three-dimensional neutronic calculations were performed using the Monte-Carlo code MCNP and the nuclear data library FENDL/2. Furthermore, the activation calculations of HCSB-TBM were carried out with the European activation system EASY-2007. At shutdown the total activity is 1.29 x 10 16 Bq, and the total afterheat is 2.46 kW. They are both dominated by the Eurofer steel. The activity and afterheat are both in the safe range of TBM design, and will not have a great impact on the environment. Meanwhile,on basis of the calculated contact dose rate, the activated materials can be re-used following the remote handling recycling options. The activation results demonstrate that the current HCSB-TBM design can satisfy the ITER safety design requirements from the activation point of view. (authors)

  15. Low-cycle fatigue of heat-resistant alloys in high-temperature gas-cooled reactor helium

    Tsuji, H.; Kondo, T.

    1984-01-01

    Strain controlled low-cycle fatigue tests were conducted on four nickel-base heat-resistant alloys at 900 0 C in simulated high-temperature gas-cooled reactor (HTGR) environments and high vacuums of about 10 -6 Pa. The observed behaviors of the materials were different and divided into two groups when tests were made in simulated HTGR helium, while all materials behaved similarly in vacuums. The materials that have relatively high ductility and compatibility with impure helium at test temperature showed considerable resistance to the fatigue damage in impure helium. On the other hand, the alloys qualified with their high creep strength were seen to suffer from the adverse effects of impure helium and the trend of intergranular cracking as well. The results were analyzed in terms of their susceptibility to the environmentenhanced fatigue damage by examining the ratios of the performance in impure helium to in vacuum. The materials that showed rather unsatisfactory resistance were considered to be characterized by their limited ductility partly due to their coarse grain structure and susceptibility to intergranular oxidation. Moderate carburization was commonly noted in all materials, particularly at the cracked portions, indicating that carbon intrusion had occurred during the crack growth stage

  16. X-ray tomography investigations on pebble bed structures

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

    2007-01-01

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

  17. Design and research on the measurement platform of the effective thermal conductivity for Li{sub 4}SiO{sub 4} and Li{sub 2}TiO{sub 3} pebble bed

    Li, Yuanjie, E-mail: yuanjli@ustc.edu.cn; Yang, Wanli; Jin, Cheng; Zhao, Pinghui; Chen, Hongli

    2015-10-15

    China is carrying out the conceptual design of Chinese Fusion Engineering Testing Reactor (CFETR), and the Helium Cooled Pebble Bed (HCPB) blanket concept is one of the main choices for tritium production. Li{sub 4}SiO{sub 4} and Li{sub 2}TiO{sub 3} are the candidate breeder materials for the HCPB blanket concept. In the HCPB blanket, breeding pebbles with the diameter range of 0.6–1.2 mm are placed between two plates and the bed shall be cooled. Accordingly, effective thermal conductivity of pebble beds needs to be determined for the heat transfer calculation. Measurements of the heat transfer parameters of Li{sub 4}SiO{sub 4} and Li{sub 2}TiO{sub 3} pebble beds are being performed at the University of Science and Technology of China (USTC). Two measurement methods are being used. One is the steady state method with the use of thermocouples to measure the temperature distribution of the pebble bed. Another is transient thermal probe method using the temperature variation of the thermal probe and Monte Carlo inversion method to calculate the heat transfer parameters of the pebble bed. This paper will report on the progress of these measurement platforms.

  18. Concept Design for a High Temperature Helium Brayton Cycle with Interstage Heating and Cooling

    Wright, Steven A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vernon, Milton E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Pickard, Paul S. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2013-12-01

    The primary metric for the viability of these next generation nuclear power plants will be the cost of generated electricity. One important component in achieving these objectives is the development of power conversion technologies that maximize the electrical power output of these advanced reactors for a given thermal power. More efficient power conversion systems can directly reduce the cost of nuclear generated electricity and therefore advanced power conversion cycle research is an important area of investigation for the Generation IV Program. Brayton cycles using inert or other gas working fluids, have the potential to take advantage of the higher outlet temperature range of Generation IV systems and allow substantial increases in nuclear power conversion efficiency, and potentially reductions in power conversion system capital costs compared to the steam Rankine cycle used in current light water reactors. For the Very High Temperature Reactor (VHTR), Helium Brayton cycles which can operate in the 900 to 950 C range have been the focus of power conversion research. Previous Generation IV studies examined several options for He Brayton cycles that could increase efficiency with acceptable capital cost implications. At these high outlet temperatures, Interstage Heating and Cooling (IHC) was shown to provide significant efficiency improvement (a few to 12%) but required increased system complexity and therefore had potential for increased costs. These scoping studies identified the potential for increased efficiency, but a more detailed analysis of the turbomachinery and heat exchanger sizes and costs was needed to determine whether this approach could be cost effective. The purpose of this study is to examine the turbomachinery and heat exchanger implications of interstage heating and cooling configurations. In general, this analysis illustrates that these engineering considerations introduce new constraints to the design of IHC systems that may require

  19. Development of local heat transfer and pressure drop models for pebble bed high temperature gas-cooled reactor cores - HTR2008-58296

    McLaughlin, B.; Worsley, M.; Stainsby, R.; Grief, A.; Dennier, A.; Macintosh, S.; Van Heerden, E.

    2008-01-01

    This paper describes pressure drop and heat transfer coefficient predictions for a typical coolant flow within the core of a pebble bed reactor (PBR) by examining a representative group of pebbles remote from the reflector region. The three- dimensional steady state flow and heat transfer predictions utilized in this work are obtained from a computational fluid dynamics (CFD) model created in the commercial software ANSYS FLUENT TM . This work utilizes three RANS turbulence models and the Chilton-Colburn analogy for heat transfer. A methodology is included in this paper for creating a quality unstructured mesh with prismatic surface layers on a random arrangement of touching pebbles. The results of the model are validated by comparing them with the correlations of the German KTA rules for a PBR. (authors)

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

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

    2016-01-01

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

  1. Pebble red modular reactor - South Africa

    Fox, M.; Mulder, E.

    1996-01-01

    In 1995 the South African Electricity Utility, ESKOM, was convinced of the economical advantages of high temperature gas-cooled reactors as viable supply side option. Subsequently planning of a techno/economic study for the year 1996 was initiated. Continuation to the construction phase of a prototype plant will depend entirely on the outcome of this study. A reactor plant of pebble bed design coupled with a direct helium cycle is perceived. The electrical output is limited to about 100 MW for reasons of safety, economics and flexibility. Design of the reactor will be based on internationally proven, available technology. An extended research and development program is not anticipated. New licensing rules and regulations will be required. Safety classification of components will be based on the merit of HTGR technology rather than attempting to adhere to traditional LWR rules. A medium term time schedule for the design and construction of a prototype plant, commissioning and performance testing is proposed during the years 2002 and 2003. Pending the performance outcome of this plant and the current power demand, series production of 100 MWe units is foreseen. (author)

  2. Thermal insulation of the high-temperature helium-cooled reactors

    Kharlamov, A.G.; Grebennik, V.N.

    1979-01-01

    Unlike the well-known thermal insulation methods, development of high-temperature helium reactors (HTGR) raises quite new problems. To understand these problems, it is necessary to consider behaviour of thermal insulation inside the helium circuit of HTGR and requirements imposed on it. Substantiation of these requirements is given in the presented paper

  3. Neutronics - thermal-hydraulics coupling: application to the helium-cooled fast reactor

    Vaiana, F.

    2009-11-01

    This thesis focuses on the study of interactions between neutron-kinetics and thermal-hydraulics. Neutron-kinetics allow to calculate the power in a nuclear reactor and the temperature evolution of materials where this power is deposited is known thanks to thermal-hydraulics. Moreover, when the temperatures evolve, the densities and cross sections change. These two disciplines are thus coupled. The first part of this work corresponds to the study and development of a method which allows to simulate transients in nuclear reactors and especially with a Monte-Carlo code for neutron-kinetics. An algorithm for the resolution of the neutron transport equation has been established and validated with a benchmark. In thermal-hydraulics, a porous media approach, based on another thesis, is considered. This gives the opportunity to solve the equations on the whole core without unconscionable computation time. Finally, a theoretical study has been performed on the statistical uncertainties which result from the use of a Monte-Carlo code and which spread from the reactivity to the power and from the power to the temperatures. The second part deals with the study of a misplaced control rod withdrawing in a GFR (helium-cooled fast reactor), a fourth generation reactor. Some models allowing to calculate neutron-kinetics and thermal-hydraulics in the core (which contains assemblies built up with fuel plates) were defined. In thermal-hydraulics, a model for the core based on the porous media approach and a fuel plate homogenization model have been set up. A similar homogenization model has been studied for neutron-kinetics. Finally, the control rod withdrawing transient where we can observe the power raising and the stabilisation by thermal feedback has been performed with the Monte-Carlo code Tripoli for neutron-kinetics and the code Trio-U for thermal-hydraulics. (author)

  4. Design and optimization of a multistage turbine for helium cooled reactor

    Braembussche, R.A. van den; Brouckaert, J.F.; Paniagua, G.; Briottet, L.

    2008-01-01

    This paper describes the aerodynamic design and explores the performance limits of a 600 MWt multistage helium turbine for a high temperature nuclear reactor closed cycle gas turbine. The design aims for maximum performance while limiting the number of stages for reasons of rotor dynamics and weight. A first part discusses the arguments that allow a preliminary selection of the overall dimensions by means of performance prediction correlations and simplified stress considerations. The rotational speed being fixed at 3000 rpm, the only degrees of freedom for the design are: the impeller diameter, number of stages and stage loading. The optimum load distribution of the different stages, the main flow parameters and the blade overall dimensions are defined by means of a 2D through-flow analysis method. The resulting absolute and relative flow angles and span-wise velocity variation are the input for a first detailed design by an inverse method. The latter defines the different 2D blade sections corresponding to prescribed optimum velocity distributions. The final 3D blade definition is made by means of a computer based 3D-DESIGN system developed at the von Karman Institute. This method combines a 3D Navier-Stokes (NS) solver, Database, Artificial Neural Network and Genetic Algorithm into a two level optimization technique for compressor and turbine stages. The use of 3D Navier-Stokes solvers allows full accounting of the secondary flow losses and optimization of the compound leaning of the stator vanes. The performance of the individual stages is used to define the multistage operating curves. The last part of the paper describes an evaluation of the cooling requirements of the first turbine rotor

  5. Numerical studies on helium cooled divertor finger mock up with sectorial extended surfaces

    Rimza, Sandeep; Satpathy, Kamalakanta; Khirwadkar, Samir; Velusamy, Karupanna

    2014-01-01

    Highlights: • Studies on heat transfer enhancement for divertor finger mock-up. • Heat transfer characteristics of jet impingement with extended surfaces have been investigated. • Effect of critical parameters that influence the thermal performance of the finger mock-up by CFD approach. • Effect of extended surface in enhancing heat removal potential with pumping power assessed. • Practicability of the chosen design is verified by structural analysis. - Abstract: Jet impinging technique is an advance divertor concept for the design of future fusion power plants. This technique is extensively used due to its high heat removal capability with reasonable pumping power and for safe operation. In this design, plasma-facing components are fabricated with numerous fingers cooled by helium jets to reduce the thermal stresses. The present study is focused towards finding an optimum performance of one such finger mock-up through systematic computational fluid dynamics (CFD) studies. Heat transfer characteristics of jet impingement have been numerically investigated with sectorial extended surfaces (SES). The result shows that addition of SES enhances heat removal potential with minimum pumping power. Detailed parametric studies on critical parameters that influence thermal performance of the finger mock-up have been analyzed. Thermo-mechanical analysis has been carried out through finite element based approach to know the state of stress in the assembly as a result of large temperature gradients. It is seen that the stresses are within the permissible limits for the present design. The whole numerical simulation has been carried out using general-purpose CFD software (ANSYS FLUENT, Release 14.0, User Guide, Ansys, Inc., 2011). Benchmark validation studies have been performed against high-heat flux experiments (B. Končar, P. Norajitra, K. Oblak, Appl. Therm. Eng., 30, 697–705, 2010) and a good agreement is noticed between the present simulation and the reported

  6. Preliminary electromagnetic analysis of Helium Cooled Solid Blanket for CFETR by MAXWELL

    Jin, Cheng; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn

    2016-11-15

    Highlights: • A FEM model of the blanket and magnetic system was built. • Electromagnetic forces and moments of the typical blanket for ferromagnetic and non-ferromagnetic materials were computed and analyzed. • Maxwell forces and Lorentz forces were computed and compared. • Eddy current in the blanket was analyzed under MD condition. - Abstract: A Helium Cooled Solid Blanket (HCSB) for CFETR (Chinese Fusion Engineering Test Reactor) was designed by USTC. The structural and thermal-hydraulic analysis has been carried out, while electromagnetic analysis was not carefully researched. In this paper, a FEM (finite element method) model of the HCSB was developed and electromagnetic forces as well as moments was computed by a FEM software called MAXWELL integrated in ANSYS Workbench. In the geometrical model, flow channels and small connecting parts were neglected because of the extreme complication and the reasonable conservative assumption by neglecting these circumstantial details. As for electromagnetic (EM) analysis, Lorentz forces due to eddy currents caused by main disruption and Maxwell forces due to the magnetization of RAFM steel (i.e. EUROFER97) were computed. Since the unavailability of the details of the plasma in CFETR, when disruptions happen, the condition where a linear current quench of main disruption occurs was assumed. The maximum magnitude of the electromagnetic forces was 356.45 kN and the maximum value of the coupled electromagnetic moments was 1899.40 N m around the radial direction. It is feasible to couple electromagnetic analysis, structural analysis and thermal-hydraulic analysis in the future since MAXWELL has good channels to exchange data between different analytic parts.

  7. Electromagnetic analysis of the Korean helium cooled ceramic reflector test blanket module set

    Lee, Youngmin; Ku, Duck Young; Lee, Dong Won; Ahn, Mu-Young; Park, Yi-Hyun; Cho, Seungyon

    2016-01-01

    Korean helium cooled ceramic reflector (HCCR) test blanket module set (TBM-set) will be installed at equatorial port #18 of Vacuum Vessel in ITER in order to test the breeding blanket performance for forthcoming fusion power plant. Since ITER tokamak has a set of electromagnetic coils (Central Solenoid, Poloidal Field and Toroidal Field coil set) around Vacuum Vessel, the HCCR TBM-set, the TBM and associated shield, is greatly influenced by magnetic field generated by these coils. In the case of fast transient electromagnetic events such as major disruption, vertical displacement event or magnet fast discharge, magnetic field and induced eddy current results in huge electromagnetic load, known as Lorentz load, on the HCCR TBM-set. In addition, the TBM-set experiences electromagnetic load due to magnetization of the structural material not only during the fast transient events but also during normal operation since the HCCR TBM adopts Reduced Activation Ferritic Martensitic (RAFM) steel as a structural material. This is known as Maxwell load which includes Lorentz load as well as load due to magnetization of structure material. This paper presents electromagnetic analysis results for the HCCR TBM-set. For analysis, a 20° sector finite model was constructed considering ITER configuration such as Vacuum Vessel, ITER shield blankets, Central Solenoid, Poloidal Field, Toroidal Field coil set as well as the HCCR TBM-set. Three major disruptions (operational event, likely event and highly unlikely event) were selected for analysis based on the load specifications. ANSYS-EMAG was used as a calculation tool. The results of EM analysis will be used as input data for the structural analysis.

  8. Electromagnetic analysis of the Korean helium cooled ceramic reflector test blanket module set

    Lee, Youngmin, E-mail: ymlee@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Ku, Duck Young [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahn, Mu-Young; Park, Yi-Hyun; Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    Korean helium cooled ceramic reflector (HCCR) test blanket module set (TBM-set) will be installed at equatorial port #18 of Vacuum Vessel in ITER in order to test the breeding blanket performance for forthcoming fusion power plant. Since ITER tokamak has a set of electromagnetic coils (Central Solenoid, Poloidal Field and Toroidal Field coil set) around Vacuum Vessel, the HCCR TBM-set, the TBM and associated shield, is greatly influenced by magnetic field generated by these coils. In the case of fast transient electromagnetic events such as major disruption, vertical displacement event or magnet fast discharge, magnetic field and induced eddy current results in huge electromagnetic load, known as Lorentz load, on the HCCR TBM-set. In addition, the TBM-set experiences electromagnetic load due to magnetization of the structural material not only during the fast transient events but also during normal operation since the HCCR TBM adopts Reduced Activation Ferritic Martensitic (RAFM) steel as a structural material. This is known as Maxwell load which includes Lorentz load as well as load due to magnetization of structure material. This paper presents electromagnetic analysis results for the HCCR TBM-set. For analysis, a 20° sector finite model was constructed considering ITER configuration such as Vacuum Vessel, ITER shield blankets, Central Solenoid, Poloidal Field, Toroidal Field coil set as well as the HCCR TBM-set. Three major disruptions (operational event, likely event and highly unlikely event) were selected for analysis based on the load specifications. ANSYS-EMAG was used as a calculation tool. The results of EM analysis will be used as input data for the structural analysis.

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

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

    2012-01-01

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

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

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

    2012-07-01

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

  11. Adsorption removal of carbon dioxide from the helium coolant of high-temperature gas-cooled reactors

    Varezhin, A.V.; Fedoseenkov, A.N.; Khrulev, A.A.; Metlik, I.V.; Zel venskii, Y.D.

    1986-01-01

    This paper conducts experiments on the removal of CO 2 from helium by means of a Soviet-made adsorbent under the conditions characteristic of high-temperature gas-cooled reactor cleaning systems. The adsorption of CO 2 from helium was studied under dynamic conditions with a fixed layer of adsorbent in a flow-through apparatus with an adsorber 16 mm in diameter. The analysis of the helium was carried out by means of a TVT chromatograph. In order to compare the adsorption of CO 2 on CaA zeolite under dynamic conditions from the helium stream under pressure with the equilibrium adsorption on the basis of pure CO 2 , the authors determined the adsorption isotherm at 293 K by the volumetric method over a range of CO 2 equilibrium pressures from 260 to 11,970 Pa. Reducing the adsorption temperature to 273 K leads to a considerable reduction in the energy costs for regeneration, owing to the increase in adsorption and the decrease in the number of regeneration cycles; the amount of the heating gas used is reduced to less than half

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

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

    2016-08-15

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

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

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

    2016-01-01

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

  14. Advanced Gas Cooled Reactor Materials Program. Reducing helium impurity depletion in HTGR materials testing

    Baldwin, D.H.

    1984-08-01

    Moisture depletion in HTGR materials testing rigs has been empirically studied in the GE High Temperature Reactor Materials Testing Laboratory (HTRMTL). Tests have shown that increased helium flow rates and reduction in reactive (oxidizable) surface area are effective means of reducing depletion. Further, a portion of the depletion has been shown to be due to the presence of free C released by the dissociation of CH 4 . This depletion component can be reduced by reducing the helium residence time (increasing the helium flow rate) or by reducing the CH 4 concentration in the test gas. Equipment modifications to reduce depletion have been developed, tested, and in most cases implemented in the HTRMTL to date. These include increasing the Helium Loop No. 1 pumping capacity, conversion of metallic retorts and radiation shields to alumina, isolation of thermocouple probes from the test gas by alumina thermowells, and substitution of non-reactive Mo-TZM for reactive metallic structural components

  15. Numerical Analysis for Heat transfer characteristic of Helium cooling system in Helium cooled ceramic reflector Test Module Blanket (HCCR-TBM)

    Park, Seong Dae; Lee, Dong Won; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae Sung; Kim, Suk Kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The main objectives of ITER project can be summarized into three types as follows - Plasma operation for a long time - Large tokamak device technology - Test blanket module (TBM) installation and verification The thermal-hydraulic analysis was performed in the He cooling channel in the BZ region of the HCCR TBM. The maximum temperature in the breeder material is equal to the limit temperature in the present design cooling channel. Nuclear fusion energy has advantage in terms of safety, resource availability, cost and waste management. There is not enough experimental results about the fusion reactor due to the severe experiments restrictions like vacuum environment, plasma production and significant nuclear heating at the same time. Much research and time is required for the commercial fusion reactor. For technical verification against the commercialization of fusion reactor, 7 countries which are EU, USA, Japan, Russia, China, India, and South Korea are building an ITER in the south of France. New designed cooling channels were proposed to improve the cooling performance. The swirl flow accelerates the mixture flow in the channels.

  16. SAFE AND FAST QUENCH RECOVERY OF LARGE SUPERCONDUCTING SOLENOIDS COOLED BY FORCED TWO-PHASE HELIUM FLOW

    Jia, L.X.

    1999-01-01

    The cryogenic characteristics in energy extraction of the four fifteen-meter-diameter superconducting solenoids of the g-2 magnet are reported in this paper. The energy extraction tests at full-current and half-current of its operating value were deliberately carried out for the quench analyses and evaluation of the cryogenic system. The temperature profiles of each coil mandrel and pressure profiles in its helium cooling tube during the energy extraction are discussed. The low peak temperature and pressure as well as the short recovery time indicated the desirable characteristics of the cryogenic system

  17. Dynamic simulation for scram of high temperature gas-cooled reactor with indirect helium turbine cycle system

    Li Wenlong; Xie Heng

    2011-01-01

    A dynamic analysis code for this system was developed after the mathematical modeling and programming of important equipment of 10 MW High Temperature Gas Cooled Reactor Helium Turbine Power Generation (HTR-10GT), such as reactor core, heat exchanger and turbine-compressor system. A scram accident caused by a 0.1 $ reactivity injection at 5 second was simulated. The results show that the design emergency shutdown plan for this system is safe and reasonable and that the design of bypass valve has a large safety margin. (authors)

  18. Engineering structure design and fabrication process of small sized China helium-cooled solid breeder test blanket module

    Wang Zeming; Chen Lu; Hu Gang

    2014-01-01

    Preliminary design and analysis for china helium-cooled solid breeder (CHHC-SB) test blanket module (TBM) have been carried out recently. As partial verification that the original size module was reasonable and the development process was feasible, fabrication work of a small sized module was to be carried out targetedly. In this paper, detailed design and structure analysis of small sized TBM was carried out based on preliminary design work, fabrication process and integrated assembly process was proposed, so a fabrication for the trial engineering of TBM was layed successfully. (authors)

  19. Adopted Methodology for Cool-Down of SST-1 Superconducting Magnet System: Operational Experience with the Helium Refrigerator

    Sahu, A. K.; Sarkar, B.; Panchal, P.; Tank, J.; Bhattacharya, R.; Panchal, R.; Tanna, V. L.; Patel, R.; Shukla, P.; Patel, J. C.; Singh, M.; Sonara, D.; Sharma, R.; Duggar, R.; Saxena, Y. C.

    2008-03-01

    The 1.3 kW at 4.5 K helium refrigerator / liquefier (HRL) was commissioned during the year 2003. The HRL was operated with its different modes as per the functional requirements of the experiments. The superconducting magnets system (SCMS) of SST-1 was successfully cooled down to 4.5 K. The actual loads were different from the originally predicted boundary conditions and an adjustment in the thermodynamic balance of the refrigerator was necessary. This led to enhanced capacity, which was achieved without any additional hardware. The required control system for the HRL was tuned to achieve the stable thermodynamic balance, while keeping the turbines' operating parameters at optimized conditions. An extra mass flow rate requirement was met by exploiting the margin available with the compressor station. The methodology adopted to modify the capacity of the HRL, the safety precautions and experience of SCMS cool down to 4.5 K, are discussed.

  20. A preliminary definition of the parameters of an experimental natural - uranium, graphite - moderated, helium - cooled power reactor

    Baltazar, O.

    1978-01-01

    A preliminary study of the technical characteristic of an experiment at 32 MWe power with a natural uconium, graphite-moderated, helium cooled reactor is described. The national participation and the use of reactor as an instrument for the technological development of future high temperature gas cooled reactor is considered in the choice of the reactor type. Considerations about nuclear power plants components based in extensive bibliography about similar english GCR reactor is presented. The main thermal, neutronic an static characteristic and in core management of the nuclear fuel is stablished. A simplified scheme of the secondary system and its thermodynamic performance is determined. A scheme of parameters calculation of the reactor type is defined based in the present capacity of calculation developed by Coordenadoria de Engenharia Nuclear and Centro de Processamento de Dados, IEA, Brazil [pt

  1. Neutronics study on hybrid reactor cooled by helium, water and molten salt

    Li Zaixin; Feng Kaiming; Zhang Guoshu; Zheng Guoyao; Zhao Fengchao

    2009-01-01

    There is no serious magnetohydrodynamics (MHD) problem when helium,water or molten salt of Flibe flows in high magnetic field. Thus helium, water and Flibe were proposed as candidate of coolant for fusion-fission hybrid reactor based on magnetic confinement. The effect on neutronics of hybrid reactor due to coolant was investigated. The analyses of neutron spectra and fuel breeding of blanket with different coolants were performed. Variations of tritium breeding ratio (TBR), blanket energy multiplication (M) and keff with operating time were also studied. MCNP code was used for neutron transport simulation. It is shown that spectra change greatly with different coolants. The blanket with helium exhibits very hard spectrum and good tritium breeding ability. And fission reactions are mainly from fast neutron. The blanket with water has soft spectrum and high energy multiplication factor. However, it needs to improve TBR. The blanket with Flibe has hard spectrum and less energy release. (authors)

  2. Adsorption purification of helium coolant of high-temperature gas-cooled reactors of carbon dioxide

    Varezhkin, A.V.; Zel'venskij, Ya.D.; Metlik, I.V.; Khrulev, A.A.; Fedoseenkin, A.N.

    1986-01-01

    A series experiments on adsorption purification of helium of CO 2 using national adsorbent under the conditions characteristic of HTGR type reactors cleanup system is performed. The experimnts have been conducted under the dynamic mode with immobile adsorbent layer (CaA zeolite) at gas flow rates from 0,02 to 0,055 m/s in the pressure range from 0,8 to 5 MPa at the temperature of 273 and 293 K. It is shown that the adsorption grows with the decrease of gas rate, i.e. with increase of contact time with adsorbent. The helium pressure, growth noticeably whereas the temperature decrease from 293 to 273 K results in adsorption 2,6 times increase. The conclusion is drawn that it is advisable drying and purification of helium of CO 2 to perform separately using different zeolites: NaA - for water. CaA - for CO 2 . Estimations of purification unit parameters are realized

  3. Study of heat transfer in superconducting cable electrical insulation of accelerator magnet cooled by superfluid helium; Etude des transferts de chaleur dans les isolations electriques de cables supraconducteurs d'aimant d'accelerateur refroidi par helium superfluide

    Baudouy, B

    1996-10-04

    Heat transfer studies of electrical cable insulation in superconducting winding are of major importance for stability studies in superconducting magnets. This work presents an experimental heat transfer study in superconducting cables of Large Hadron Collider dipoles cooled by superfluid helium and submitted to volume heat dissipation due to beam losses. For NbTi magnets cooled by superfluid helium the most severe heat barrier comes from the electrical insulation of the cables. Heat behaviour of a winding is approached through an experimental model in which insulation characteristics can be modified. Different tests on insulation patterns show that heat transfer is influenced by superfluid helium contained in insulation even for small volume of helium (2 % of cable volume). Electrical insulation can be considered as a composite material made of a solid matrix with a helium channels network which cannot be modelled easily. This network is characterised by another experimental apparatus which allows to study transverse and steady-state heat transfer through an elementary insulation pattern. Measurements in Landau regime ({delta}T{approx}10{sup -5} to 10{sup -3} K) and in Gorter-Mellink regime ({delta}T>10{sup -3} K) and using assumptions that helium thermal paths and conduction in the insulation are decoupled allow to determine an equivalent channel area (10{sup -6} m{sup 2}) and an equivalent channel diameter (25 {mu}). (author)

  4. Study of heat transfer in superconducting cable electrical insulation of accelerator magnet cooled by superfluid helium; Etude des transferts de chaleur dans les isolations electriques de cables supraconducteurs d'aimant d'accelerateur refroidi par helium superfluide

    Baudouy, B

    1996-10-04

    Heat transfer studies of electrical cable insulation in superconducting winding are of major importance for stability studies in superconducting magnets. This work presents an experimental heat transfer study in superconducting cables of Large Hadron Collider dipoles cooled by superfluid helium and submitted to volume heat dissipation due to beam losses. For NbTi magnets cooled by superfluid helium the most severe heat barrier comes from the electrical insulation of the cables. Heat behaviour of a winding is approached through an experimental model in which insulation characteristics can be modified. Different tests on insulation patterns show that heat transfer is influenced by superfluid helium contained in insulation even for small volume of helium (2 % of cable volume). Electrical insulation can be considered as a composite material made of a solid matrix with a helium channels network which cannot be modelled easily. This network is characterised by another experimental apparatus which allows to study transverse and steady-state heat transfer through an elementary insulation pattern. Measurements in Landau regime ({delta}T{approx}10{sup -5} to 10{sup -3} K) and in Gorter-Mellink regime ({delta}T>10{sup -3} K) and using assumptions that helium thermal paths and conduction in the insulation are decoupled allow to determine an equivalent channel area (10{sup -6} m{sup 2}) and an equivalent channel diameter (25 {mu}). (author)

  5. An efficient cooling loop for connecting cryocooler to a helium reservoir

    Taylor, C.E.; Abbott, C.S.R.; Leitner, D.; Leitner, M.; Lyneis, C.M.

    2003-01-01

    The magnet system of the VENUS ECR Ion Source at LBNL has two 1.5-watt cryocoolers suspended in the cryostat vacuum. Helium vapor from the liquid reservoir is admitted to a finned condenser bolted to the cryocooler 2nd stage and returns as liquid via gravity. Small-diameter flexible tubes allow the cryocoolers to be located remotely from the reservoir. With 3.1 watts load, the helium reservoir is maintained at 4.35 K, 0.05K above the cryocooler temperature. Design, analysis, and performance are presented

  6. Evaluation, Comparison and Optimization of the Compact Recuperator for the High Temperature Gas-Cooled Reactor (HTGR) Helium Turbine System

    Hao Haoran; Yang Xiaoyong; Wang Jie; Ye Ping; Yu Xiaoli; Zhao Gang

    2014-01-01

    Helium turbine system is a promising method to covert the nuclear power generated by the High Temperature Gas Cooled Reactor (HTGR) into electricity with inherent safety, compact configuration and relative high efficiency. And the recuperator is one of the key components for the HTGR helium turbine system. It is used to recover the exhaust heat out of turbine and pass it to the helium from high pressure compressor, and hence increase the cycle’s efficiency dramatically. On the other hand, the pressure drop within the recuperator will reduce the cycle efficiency, especially on low pressure side of recuperator. It is necessary to optimize the design of recuperator to achieve better performance of HTGR helium turbine system. However, this optimization has to be performed with the restriction of the size of the pressure vessel which contains the power conversion unit. This paper firstly presents an analysis to investigate the effects of flow channel geometry, recuperator’s power and size on heat transfer and pressure drop. Then the relationship between the recuperator design and system performance is established with an analytical model, followed by the evaluations of the current recuperator designs of GT-MHR, GTHTR300 and PBMR, in which several effective technical measures to optimize the recuperator are compared. Finally it is found that the most important factors for optimizing recuperator design, i.e. the cross section dimensions and tortuosity of flow channel, which can also be extended to compact intermediate heat exchangers. It turns out that a proper optimization can increase the cycle’s efficiency by 1~2 percentage, which could also raise the economy and competitiveness of future commercial HTGR plants. (author)

  7. Initial assessment of environmental effects on SiC/SiC composites in helium-cooled nuclear systems

    Contescu, Cristian I [ORNL

    2013-09-01

    This report summarized the information available in the literature on the chemical reactivity of SiC/SiC composites and of their components in contact with the helium coolant used in HTGR, VHTR and GFR designs. In normal operation conditions, ultra-high purity helium will have chemically controlled impurities (water, oxygen, carbon dioxide, carbon monoxide, methane, hydrogen) that will create a slightly oxidizing gas environment. Little is known from direct experiments on the reactivity of third generation (nuclear grade) SiC/SiC composites in contact with low concentrations of water or oxygen in inert gas, at high temperature. However, there is ample information about the oxidation in dry and moist air of SiC/SiC composites at high temperatures. This information is reviewed first in the next chapters. The emphasis is places on the improvement in material oxidation, thermal, and mechanical properties during three stages of development of SiC fibers and at least two stages of development of the fiber/matrix interphase. The chemical stability of SiC/SiC composites in contact with oxygen or steam at temperatures that may develop in off-normal reactor conditions supports the conclusion that most advanced composites (also known as nuclear grade SiC/SiC composites) have the chemical resistance that would allow them maintain mechanical properties at temperatures up to 1200 1300 oC in the extreme conditions of an air or water ingress accident scenario. Further research is needed to assess the long-term stability of advanced SiC/SiC composites in inert gas (helium) in presence of very low concentrations (traces) of water and oxygen at the temperatures of normal operation of helium-cooled reactors. Another aspect that needs to be investigated is the effect of fast neutron irradiation on the oxidation stability of advanced SiC/SiC composites in normal operation conditions.

  8. Thermodynamic analysis and economical evaluation of two 310-80 K pre-cooling stage configurations for helium refrigeration and liquefaction cycle

    Zhu, Z. G.; Zhuang, M.; Jiang, Q. F.; Y Zhang, Q.; Feng, H. S.

    2017-12-01

    In 310-80 K pre-cooling stage, the temperature of the HP helium stream reduces to about 80 K where nearly 73% of the enthalpy drop from room temperature to 4.5 K occurs. Apart from the most common liquid nitrogen pre-cooling, another 310-80 K pre-cooling configuration with turbine is employed in some helium cryoplants. In this paper, thermodynamic and economical performance of these two kinds of 310-80 K pre-cooling stage configurations has been studied at different operating conditions taking discharge pressure, isentropic efficiency of turbines and liquefaction rate as independent parameters. The exergy efficiency, total UA of heat exchangers and operating cost of two configurations are computed. This work will provide a reference for choosing 310-80 K pre-cooling stage configuration during design.

  9. Design and study of Engineering Test Facility - Helium Circulator

    Jiang Huijing; Ye Ping; Zhao Gang; Geng Yinan; Wang Jie

    2015-01-01

    Helium circulator is one of the key equipment of High-temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM). In order to simulate most normal and accident operating conditions of helium circulator in HTR-PM, a full scale, rated flow rate and power, engineering test loop, which was called Engineering Test Facility - Helium Circulator (ETF-HC), was designed and established. Two prototypes of helium circulator, which was supported by Active Magnetic Bearing (AMB) or sealed by dry gas seals, would be tested on ETF-HC. Therefore, special interchangeable design was under consideration. ETF-HC was constructed compactly, which consisted of eleven sub-systems. In order to reduce the flow resistance of the circuit, special ducts, elbows, valves and flowmeters were selected. Two stages of heat exchange loops were designed and a helium - high pressure pure water heat exchanger was applied to ensure water wouldn't be vaporized while simulating accident conditions. Commissioning tests were carried out and operation results showed that ETF-HC meets the requirement of helium circulator operation. On this test facility, different kinds of experiments were supposed to be held, including mechanical and aerodynamic performance tests, durability tests and so on. These tests would provide the features and performance of helium circulator and verify its feasibility, availability and reliability. (author)

  10. Evaluation of materials for heat exchanging components in advanced helium-cooled reactors

    Schubert, F.

    1984-01-01

    The qualification of metallic materials for advanced HTR applications is based on creep behaviour, fatigue properties, structural stability and corrosion resistance. A brief state of the art is provided for the materials for heat exchanging components. The experimental results are treated with respect to the importance for the design, the characteristic of time-depend materials behaviour are evaluated. Of specific interest are the possible effects of helium on the mechanical properties. Helium, which serves as primary coolant, contains traces of reactive impurities such as hydrogen, methane, carbon monoxide and water vapor. The evaluation of the HTR materials program serves as basis for structural design rules of components with operation temperatures above 800 deg C. The materials mechanical topics are discussed. Alloy improvement and the progress in development of new alloys are reviewed. (author)

  11. The real gas behaviour of helium as a cooling medium for high-temperature reactors

    Hewing, G.

    1977-01-01

    The article describes the influence of the real gas behaviour on the variables of state for the helium gas and the effects on the design of high-temperature reactor plants. After explaining the basic equations for describing variables and changes of state of the real gas, the real and ideal gas behaviour is analysed. Finally, the influence of the real gas behaviour on the design of high-temperature reactors in one- and two-cycle plants is investigated. (orig.) [de

  12. Critical Current and Stability of MgB$_2$ Twisted-Pair DC Cable Assembly Cooled by Helium Gas

    AUTHOR|(CDS)2069632; Ballarino, Amalia; Yang, Yifeng; Young, Edward Andrew; Bailey, Wendell; Beduz, Carlo

    2013-01-01

    Long length superconducting cables/bus-bars cooled by cryogenic gases such as helium operating over a wider temperature range are a challenging but exciting technical development prospects, with applications ranging from super-grid transmission to future accelerator systems. With limited existing knowledge and previous experiences, the cryogenic stability and quench protection of such cables are crucial research areas because the heat transfer is reduced and temperature gradient increased compared to liquid cryogen cooled cables. V-I measurements on gas-cooled cables over a significant length are an essential step towards a fully cryogenic stabilized cable with adequate quench protection. Prototype twisted-pair cables using high-temperature superconductor and MgB2 tapes have been under development at CERN within the FP7 EuCARD project. Experimental studies have been carried out on a 5-m-long multiple MgB$_2$ cable assembly at different temperatures between 20 and 30 K. The subcables of the assembly showed sim...

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

    Zhao, Xiang; Montgomery, Trent; Zhang, Sijun

    2015-01-01

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

  14. Low-temperature dynamic nuclear polarization with helium-cooled samples and nitrogen-driven magic-angle spinning.

    Thurber, Kent; Tycko, Robert

    2016-03-01

    We describe novel instrumentation for low-temperature solid state nuclear magnetic resonance (NMR) with dynamic nuclear polarization (DNP) and magic-angle spinning (MAS), focusing on aspects of this instrumentation that have not been described in detail in previous publications. We characterize the performance of an extended interaction oscillator (EIO) microwave source, operating near 264 GHz with 1.5 W output power, which we use in conjunction with a quasi-optical microwave polarizing system and a MAS NMR probe that employs liquid helium for sample cooling and nitrogen gas for sample spinning. Enhancement factors for cross-polarized (13)C NMR signals in the 100-200 range are demonstrated with DNP at 25K. The dependences of signal amplitudes on sample temperature, as well as microwave power, polarization, and frequency, are presented. We show that sample temperatures below 30K can be achieved with helium consumption rates below 1.3 l/h. To illustrate potential applications of this instrumentation in structural studies of biochemical systems, we compare results from low-temperature DNP experiments on a calmodulin-binding peptide in its free and bound states. Published by Elsevier Inc.

  15. Preliminary study on application of Pd composite membrane in helium purification system of high-temperature gas-cooled reactor

    Cai Jianhua; Yang Xiaoyong; Wang Jie; Yu Suyuan

    2008-01-01

    Helium purification system (HPS) is the main part of the helium auxiliary system of high-temperature gas-cooled reactors (HTGR), also in fusion reactors. Some exploratory work was carried out on the application of Pd composite membrane in the separation of He and H 2 . A typical single stripper permeator with recycle (SSP) system was designed, based on the design parameters of a small scale He purification test system CIGNE in CADARACHE, CEA, France, and finite element analysis method was used to solve the model. The total length of membrane module is fixed to 0.5 m. The results show that the concentration of H 2 is found to reduce from 1 000 μL/L in feed gas to 5 μL/L in the product He (the upper limitation of HPS in HTGR). And the molar ratio of product He to feed gas is 96.18% with the optimized ratio of sweep gas to retentive gas 0. 3970. It's an exponential distribution of H 2 concentration along the membrane module. The results were also compared with the other two popular designs, two stripper in series permeator (TSSP) and continuous membrane column (CMC). (authors)

  16. Mechanical characterization of metallic materials for high-temperature gas-cooled reactors in air and in helium environments

    Sainfort, G.; Cappelaere, M.; Gregoire, J.; Sannier, J.

    1984-01-01

    In the French R and D program for high-temperature gas-cooled reactors (HTGRs), three metallic alloys were studied: steel Chromesco-3 with 2.25% chromium, alloy 800H, and Hastelloy-X. The Chromesco-3 and alloy 800H creep behavior is the same in air and in HTGR atmosphere (helium). The tensile tests of Hastelloy-X specimens reveal that aging has embrittlement and hardening effects up to 700 0 C, but the creep tests at 800 0 C show opposite effects. This particular behavior could be due to induced precipitation by aging and the depletion of hardening elements from the matrix. Tests show a low influence of cobalt content on mechanical properties of Hastelloy-X

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

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

    2012-01-01

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

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

    Zhao, Xiang; Montgomery, Trent; Zhang, Sijun

    2011-01-01

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

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

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

    2018-02-01

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

  20. Pebble-bed pebble motion: Simulation and Applications

    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

  1. Pebble-bed pebble motion: Simulation and Applications

    Cogliati, Joshua J.; Ougouag, Abderrafi M.

    2011-01-01

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

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

    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

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

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

    2013-10-15

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

  4. Peculiarities of void fraction measurement applied to physical installation channels cooled by forced helium flow

    Danilov, V.V.; Filippov, Yu.P.; Mamedov, I.S.

    1989-01-01

    The methods of optimizing the transducers designed for measurements of the void fraction of two-phase flows in the channels of round and annular cross section are presented. On the basis of the analysis performed concrete solution of relatively high technical characteristics are proposed. Rated and actual characteristics of signal ranges and measurement errors are given for both sensors. Influence of the mass velocity on the void fraction of adiabatic two-phase flows is theoretically analyzed. Effects of friction and of liquid-into-vapour entrainment are shown. Calculation results are compared with the obtained experimental data for helium. Special attention is given to the specific features of the processes in channels with different cross section. 17 refs.; 5 figs.; 1 tab

  5. Cryogenic filter method produces super-pure helium and helium isotopes

    Hildebrandt, A. F.

    1964-01-01

    Helium is purified when cooled in a low pressure environment until it becomes superfluid. The liquid helium is then filtered through iron oxide particles. Heating, cooling and filtering processes continue until the purified liquid helium is heated to a gas.

  6. Pebble breakage in gravel

    Tuitz, C.

    2012-01-01

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

  7. Pebble-bed reactor

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

    1976-01-01

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

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

    Peterson, Per F.

    2013-05-14

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

  9. Numerical examination of temperature control in helium-cooled high flux test module of IFMIF

    Ebara, Shinji; Yokomine, Takehiko; Shimizu, Akihiko

    2007-01-01

    For long term irradiation of the International Fusion Materials Irradiation Facility (IFMIF), test specimens are needed to retain constant temperature to avoid change of its irradiation characteristics. The constant temperatures control is one of the most challenging issues for the IFMIF test facilities. We have proposed a new concept of test module which is capable of precisely measuring temperature, keeping uniform temperature with enhanced cooling performance. In the system according to the new design, cooling performances and temperature distributions of specimens were examined numerically under diverse conditions. Some transient behaviors corresponding to the prescribed temperature control mode were perseveringly simulated. It was confirmed that the thermal characteristics of the new design satisfied the severe requirement of IFMIF

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

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

    2018-03-01

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

  11. Sockets and Pebbles

    1997-01-01

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

  12. There is no explosion risk associated with superfluid Helium in the LHC cooling system

    Fairbairn, Malcolm

    2008-01-01

    We evaluate speculation about the possibility of a dangerous release of energy within the liquid Helium of the Large Hadron Collider (LHC) cryogenic system due to the occurrence of a "Bose-Nova". Bose-Novae are radial bursts of rapidly moving atoms which can occur when a Bose-Einstein Condensate (BEC) undergoes a collapse due the interatomic potential being deliberately made attractive using a magnetic field close to the Feshbach resonance. Liquid 4He has a monatomic structure with s-wave electrons, zero nuclear spin, no hyperfine splitting, and as a consequence no Feshbach resonance which would allow one to change its normally repulsive interactions to be attractive. Because of this, a Bose-Nova style collapse of 4He is impossible. Additional speculations concerning cold fusion during these events are easily dismissed using the usual arguments about the Coulomb barrier at low temperatures, and are not needed to explain the Bose-Einstein condensate Bose-Nova phenomenon. We conclude that that there is no physi...

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

    Chen, Jianwei

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

  14. Design and performance study of the helium-cooled T-tube divertor concept

    Ihli, T.; Raffray, A.R.; Abdel-Khalik, S.I.; Shin, S.

    2007-01-01

    The ARIES-CS study has been launched with the goal of developing through physics and engineering optimization an attractive power plant concept based on a compact stellarator configuration. The study included an effort to characterize the divertor location and corresponding heat load distribution, and to develop a He-cooled divertor concept that could accommodate a heat flux of at least 10 MW/m 2 , and that would integrate well with the other power core components. This paper describes the design study of this divertor concept, which, although developed for a compact stellarator, is well suited for a tokamak configuration also

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

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2006-10-01

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

  16. Optimization of MOX fuel cycles in pebble bed HTGR

    Wei Jinfeng; Li Fu; Sun Yuliang

    2013-01-01

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

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

    Kemeny, L.G.

    2003-01-01

    This paper describes a joint project of Massachusetts Institute of technology, Nu-Tec Inc. and Proto Power. The elegant simplicity of graphite moderated pebble bed reactor is the basis for the 'generation four' nuclear power plants. High Temperature Gas Cooled (HTGC) nuclear power plant have the potential to become the preferred base load sustainable energy source for the new millennium. The great attraction of these helium cooled 'Generation Four' nuclear plant can be summarised as follows: Factory assembly line production; Modularity and ease of delivery to site; High temperature Brayton Cycle ideally suited for cogeneration of electricity, potable water and hydrogen; Capital and operating costs competitive with hydrocarbon plant; Design is inherently meltdown proof and proliferation resistant

  18. Hypothetical accident scenario analyses for a 250-MW(t) modular high temperature gas-cooled reactor

    Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

    1985-11-01

    This paper describes calculations performed to characterize the inherent safety of a 250-MW(t), 100-MW(e), pebble bed modular high temperature gas-cooled reactor (HTGR) design with vertical in-line arrangement (i.e., upflow core with steam generators directly above the core). A variety of postulated accident sequences involving combinations of loss of forced primary coolant (helium) circulation, loss of primary coolant pressurization, and loss of heat sink were studied and were discussed

  19. System code assessment with thermal-hydraulic experiment to develop helium cooled breeding blanket for nuclear fusion reactor

    Yum, S. B.; Park, I. W.; Park, G. C.; Lee, D. W.

    2012-01-01

    By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He Cooled Molten Lithium (HCML) Test Blanket Module (TBM) for testing in the International Thermonuclear Experimental Reactor (ITER). A performance analysis for the thermal-hydraulics and a safety analysis for an accident caused by a loss of coolant for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs Multicomponent Mixture Analysis), which was developed by the Gas Cooled Reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM First Wall (FW) mock-up made from the same material as tho KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 11, 19, and 29 bar, and under various ranges of flow rate from 0.63 to 2.44kg/min with a constant wall temperature condition. In the present study, a thermal-hydraulic test was performed with the newly constructed helium supplying system, In which the design pressure and temperature were 9 MPa and 500 .deg. C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 8 MPa pressure, 0.2 kg/s flow rate, which are almost same conditions of the KO TBM FW. One-side of the mock-up was heated with a constant heat flux of 0.5 MW/m 2 using a graphite heating system, KoHLT-2 (Korea Heat Load Test Facility-2). The wall temperatures were measured using installed thermocouples, and they show a strong parity with the code results simulated under the same test conditions

  20. Experimental study on cryogenic adsorption of methane by activated carbon for helium coolant purification of High-Temperature Gas-cooled Reactor

    Chang, Hua; Wu, Zong-Xin; Jia, Hai-Jun

    2017-01-01

    Highlights: • The cryogenic CH 4 adsorption on activated carbon was studied for design of HTGR. • The breakthrough curves at different conditions were analyzed by the MTZ model. • The CH 4 adsorption isotherm was fitted well by the Toth model and the D-R model. • The work provides valuable reference data for helium coolant purification of HTGR. - Abstract: The cryogenic adsorption behavior of methane on activated carbon was investigated for helium coolant purification of high-temperature gas-cooled reactor by using dynamic column breakthrough method. With helium as carrier gas, experiments were performed at −196 °C and low methane partial pressure range of 0–120 Pa. The breakthrough curves at different superficial velocities and different feed concentrations were measured and analyzed by the mass-transfer zone model. The methane single-component adsorption isotherm was obtained and fitted well by the Toth model and the Dubinin-Radushkevich model. The adsorption heat of methane on activated carbon was estimated. The cryogenic adsorption process of methane on activated carbon has been verified to be effective for helium coolant purification of high-temperature gas-cooled reactor.

  1. Pebble Puzzle Solved

    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.

  2. A constitutive model for the thermo-mechanical behaviour of fusion-relevant pebble beds and its application to the simulation of HELICA mock-up experimental results

    Vella, G.; Maio, P.A. Di; Giammusso, R.; Tincani, A.; Orco, G. Dell

    2006-01-01

    Within the framework of the activities promoted by European Fusion Development Agreement on the technology of the Helium Cooled Pebble Bed Test Blanket Module to be irradiated in one of the ITER equatorial ports, attention has been focused on the theoretical modelling of the thermo-mechanical constitutive behaviour of both beryllium and lithiated ceramics pebble beds, that are envisaged to act respectively as neutron multiplier and tritium breeder. The thermo-mechanical behaviour of the pebble beds and their nuclear performances in terms of tritium production depend on the reactor relevant conditions (heat flux and neutron wall load), the pebble sizes and the breeder cell geometries (bed thickness, pebble packing factor, bed overall thermal conductivity). ENEA-Brasimone and the Department of Nuclear Engineering (DIN) of the Palermo University have performed intense research activities intended to investigate fusion-relevant pebble bed thermo-mechanical behaviour by adopting both experimental and theoretical approaches. In particular, ENEA has carried out several experimental campaigns on small scale mock-ups tested in out-of-pile conditions, while DIN has developed a proper constitutive model that has been implemented on commercial FEM code, for the prediction of the thermal and mechanical performances of fusion-relevant pebble beds and for the comparison with the experimental results of the ENEA tests. In that framework, HELICA mock-up has been set-up and tested to investigate the behaviour of pebble bed in reactor-relevant geometries, providing useful data sets to be numerically reproduced by means of the DIN constitutive model, contributing to its assessment. The paper presents the constitutive model developed and the main experimental results of two test campaigns on HELICA mock-up carried out at HE-FUS 3 facility of ENEA Brasimone, the geometry of the mock-up, the adopted thermal and mechanical boundary conditions and the test operating conditions. The most

  3. Cooling of interstellar formaldehyde by collision with helium: an accurate quantum mechanical calculation

    Garrison, B.J.

    1975-08-01

    In order to test a collisional pumping model as a mechanism for cooling the 6 cm and 2 cm doublets of interstellar formaldehyde, a quantum mechanical scattering calculation is performed. To obtain the intermolecular interaction between H 2 CO( 1 A 1 ) and He( 1 S) two calculations are performed, a Hartree-Fock (HF) potential surface and a configuration interaction (CI) surface. A basis set of better than ''triple zeta plus polarization'' quality is used to compute the HF portion of the potential energy surface. This portion is highly anisotropic and has a slight attraction arising from induction effects at intermolecular separations around 9 a.u. The HF surface is modified through a series of CI calculations. Correlation is found to have little effect in the strongly anisotropic repulsive region of the interaction potential but dominates the well and long-range regions. The maximum well depth is attained for in-plane approaches of He and lies in the range 35-40 0 K for arbitrary theta at center of mass separation of 7.5 a.u. The entire surface is fit to a spherical harmonic expansion to facilitate scattering applications. (auth)

  4. Potential applications of helium-cooled high-temperature reactors to process heat use

    Gambill, W.R.; Kasten, P.R.

    1981-01-01

    High-Temperature Gas-Cooled Reactors (HTRs) permit nuclear energy to be applied to a number of processes presently utilizing fossil fuels. Promising applications of HTRs involve cogeneration, thermal energy transport using molten salt systems, steam reforming of methane for production of chemicals, coal and oil shale liquefaction or gasification, and - in the longer term - energy transport using a chemical heat pipe. Further, HTRs might be used in the more distant future as the energy source for thermochemical hydrogen production from water. Preliminary results of ongoing studies indicate that the potential market for Process Heat HTRs by the year 2020 is about 150 to 250 GW(t) for process heat/cogeneration application, plus approximately 150 to 300 GW(t) for application to fossil conversion processes. HTR cogeneration plants appear attractive in the near term for new industrial plants using large amounts of process heat, possibly for present industrial plants in conjunction with molten-salt energy distribution systems, and also for some fossil conversion processes. HTR reformer systems will take longer to develop, but are applicable to chemicals production, a larger number of fossil conversion processes, and to chemical heat pipes

  5. A study on nuclear heat load tolerable for NET/TF coils cooled by internal flow of helium II

    Hofmann, A.

    1988-02-01

    NbTi cables cooled by internal flow of superfluid helium are considered an option for the design of NET/TF coils with about 11 T peak fields. Starting from an available winding cross section of 0.61x0.61 m 2 for a 8 MA turns coil made of a 16 kA conductor it is shown that sufficient hydraulic cross section can be provided within such cables to remove the expected thermal load resulting from nuclear heating with exponential decay from inboard to outboard side of the winding. The concept is a pancake type coil with 1.8 K helium fed-in the high field region of each pancake. The temperature distribution within such coils is calculated, and the local safety margin is determined from temperature and field. The calculation takes account of nuclear and a.c. heating, and of thermal conductance between the individual layers and the coil casing. It is shown that operation with 1.8 K inlet and about 3 K outlet temperature is possible. The electrical insulation with about 0.5 mm thickness proves to provide sufficient thermal insulation. No additional thermal shield is required between the coil casing and the winding package. Two different types of conductors are being considered: a) POLO type cable with quadratic cross section and a central circular coolant duct, and b) an LCT type cable with two conductors wound in hand. Both concepts with about 500 m length of the cooland channels are shown to meet the requirements resulting from a peak nuclear heat load of 0.3 mW/cm 3 in the inboard turns. The hydraulic diameters are sufficient to operate each coils with self-sustained fountain effect pumps. Even appreciably higher heat loads with up to 3 mW/cm 3 of nuclear heating can be tolerated for the POLO type cable when the hydraulic diameter is enlarged to its maximum of 17 mm. (orig.) [de

  6. Gas-cooled reactors

    Schulten, R.; Trauger, D.B.

    1976-01-01

    Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing

  7. Assessing the feasibility of a high-temperature, helium-cooled vacuum vessel and first wall for the Vulcan tokamak conceptual design

    Barnard, H.S.; Hartwig, Z.S.; Olynyk, G.M.; Payne, J.E.

    2012-01-01

    The Vulcan conceptual design (R = 1.2 m, a = 0.3 m, B 0 = 7 T), a compact, steady-state tokamak for plasma–material interaction (PMI) science, must incorporate a vacuum vessel capable of operating at 1000 K in order to replicate the temperature-dependent physical chemistry that will govern PMI in a reactor. In addition, the Vulcan divertor must be capable of handling steady-state heat fluxes up to 10 MW m −2 so that integrated materials testing can be performed under reactor-relevant conditions. A conceptual design scoping study has been performed to assess the challenges involved in achieving such a configuration. The Vulcan vacuum system comprises an inner, primary vacuum vessel that is thermally and mechanically isolated from the outer, secondary vacuum vessel by a 10 cm vacuum gap. The thermal isolation minimizes heat conduction between the high-temperature helium-cooled primary vessel and the water-cooled secondary vessel. The mechanical isolation allows for thermal expansion and enables vertical removal of the primary vessel for maintenance or replacement. Access to the primary vessel for diagnostics, lower hybrid waveguides, and helium coolant is achieved through ∼1 m long intra-vessel pipes to minimize temperature gradients and is shown to be commensurate with the available port space in Vulcan. The isolated primary vacuum vessel is shown to be mechanically feasible and robust to plasma disruptions with analytic calculations and finite element analyses. Heat removal in the first wall and divertor, coupled with the ability to perform in situ maintenance and replacement of divertor components for scientific purposes, is achieved by combining existing helium-cooled techniques with innovative mechanical attachments of plasma facing components, either in plate-type helium-cooled modules or independently bolted, helium-jet impingement-cooled tiles. The vacuum vessel and first wall design enables a wide range of potential PFC materials and configurations to

  8. Supercritical CO2 Brayton power cycles for DEMO fusion reactor based on Helium Cooled Lithium Lead blanket

    Linares, José Ignacio; Herranz, Luis Enrique; Fernández, Iván; Cantizano, Alexis; Moratilla, Beatriz Yolanda

    2015-01-01

    Fusion energy is one of the most promising solutions to the world energy supply. This paper presents an exploratory analysis of the suitability of supercritical CO 2 Brayton power cycles (S-CO 2 ) for low-temperature divertor fusion reactors cooled by helium (as defined by EFDA). Integration of three thermal sources (i.e., blanket, divertor and vacuum vessel) has been studied through proposing and analyzing a number of alternative layouts, achieving an improvement on power production higher than 5% over the baseline case, which entails to a gross efficiency (before self-consumptions) higher than 42%. In spite of this achievement, the assessment of power consumption for the circulating heat transfer fluids results in a penalty of 20% in the electricity production. Once the most suitable layout has been selected an optimization process has been conducted to adjust the key parameters to balance performance and size, achieving an electrical efficiency (electricity without taking into account auxiliary consumptions due to operation of the fusion reactor) higher than 33% and a reduction in overall size of heat exchangers of 1/3. Some relevant conclusions can be drawn from the present work: the potential of S-CO 2 cycles as suitable converters of thermal energy to power in fusion reactors; the significance of a suitable integration of thermal sources to maximize power output; the high penalty of pumping power; and the convenience of identifying the key components of the layout as a way to optimize the whole cycle performance. - Highlights: • Supercritical CO 2 Brayton cycles have been proposed for BoP of HCLL fusion reactor. • Low temperature sources have been successfully integrated with high temperature ones. • Optimization of thermal sources integration improves 5% the electricity production. • Assessment of pumping power with sources and sink loops results on 20% of gross power. • Matching of key parameters has conducted to 1/3 of reduction in heat

  9. Corrosion of high temperature alloys in the primary circuit helium of high temperature gas cooled reactors. Pt. 2

    Quadakkers, W.J.

    1985-01-01

    The reactive impurities H 2 O, CO, H 2 and CH 4 which are present in the primary coolant helium of high temperature gas-cooled reactors can cause scale formation, internal oxidation and carburization or decarburization of the high temperature structural alloys. In Part 1 of this contribution a theoretical model was presented, which allows the explanation and prediction of the observed corrosion effects. The model is based on a classical stability diagram for chromium, modified to account for deviations from equilibrium conditions caused by kinetic factors. In this paper it is shown how a stability diagram for a commercial alloy can be constructed and how this can be used to correlate the corrosion results with the main experimental parameters, temperature, gas and alloy composition. Using the theoretical model and the presented experimental results, conditions are derived under which a protective chromia based surface scale will be formed which prevents a rapid transfer of carbon between alloy and gas atmosphere. It is shown that this protective surface oxide can only be formed if the carbon monoxide pressure in the gas exceeds a critical value. Psub(CO), which depends on temperature and alloy composition. Additions of methane only have a limited effect provided that the methane/water ratio is not near to, or greater than, a critical value of around 100/1. The influence of minor alloying additions of strong oxide forming elements, commonly present in high temperature alloys, on the protective properties of the chromia surface scales and the kinetics of carbon transfer is illustrated. (orig.) [de

  10. Assessment of tritiated activities in the radwaste generated from ITER Chinese helium cooled ceramic breeding test blanket module system

    Chen, Chang An, E-mail: chenchangan@caep.cn; Liu, Lingbo; Wang, Bo; Xiang, Xin; Yao, Yong; Song, Jiangfeng

    2016-11-15

    Highlights: • Approaches were developed for calculation/evaluation of tritium activities in the materials and components of a TBM system, with tritium permeation being considered for the first time. • Almost all tritiated materials and components were considered in CNHCCB TBM system including the TBM set, connection pipes, and the ancillary tritium handling systems. • Tritium activity data in HCCB TBM system were updated. Some of which in directly tritium contacted components are to be 2 or 4 magnitudes higher than the original neutron transmutation calculations. • The radwaste amount from both operation and decommission of HCCB TBM system was evaluated. - Abstract: Chinese Helium Cooled Ceramic Breeding Test blanket Module (CNHCCB TBM) will be tested in the ITER machine for the feasibility of in pile tritium production for a future magnetic confinement fusion reactor. The tritium inventories/retentions in the material/components were evaluated and updated mainly based on the tritium diffusion/permeation theory and the analysis of some reported data. Tritiated activities rank from less than 10 Bq g{sup −1} to 10{sup 9} Bq g{sup −1} for the different materials or components, which are generally higher than those from the previous neutron transmutation calculation. The amounts of tritiated radwaste were also estimated according to the operation, decommission, maintenance and replacement strategies, which vary from several tens of kilograms to tons in the different operation phases. The data can be used both for the tritium radiological safety evaluation and radwaste management of CNHCCB TBM set and its ancillary systems.

  11. Buffer-gas cooling of antiprotonic helium to 1.5 to 1.7 K, and antiproton-to–electron mass ratio

    Hori, Masaki; Sótér, Anna; Barna, Daniel; Dax, Andreas; Hayano, Ryugo; Kobayashi, Takumi; Murakami, Yohei; Todoroki, Koichi; Yamada, Hiroyuki; Horváth, Dezső; Venturelli, Luca

    2016-01-01

    Charge, parity, and time reversal (CPT) symmetry implies that a particle and its antiparticle have the same mass. The antiproton-to-electron mass ratio Embedded Image can be precisely determined from the single-photon transition frequencies of antiprotonic helium. We measured 13 such frequencies with laser spectroscopy to a fractional precision of 2.5 × 10−9 to 16 × 10−9. About 2 × 109 antiprotonic helium atoms were cooled to temperatures between 1.5 and 1.7 kelvin by using buffer-gas cooling in cryogenic low-pressure helium gas; the narrow thermal distribution led to the observation of sharp spectral lines of small thermal Doppler width. The deviation between the experimental frequencies and the results of three-body quantum electrodynamics calculations was reduced by a factor of 1.4 to 10 compared with previous single-photon experiments. From this, Embedded Image was determined as 1836.1526734(15), which agrees with a recent proton-to-electron experimental value within 8 × 10−10.

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

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

    2013-10-01

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

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

    Prasser, Horst-Michael

    2014-01-01

    Switzerland is member of the Generation IV International Forum (GIF). The related work takes entirely place at PSI in the working groups of Gas-Cooled Fast Reactors and Very High Temperature Reactors. In the past, PSI has performed experimental and theoretical studies on criticality issues of pebble beds at the PROTEUS reactor, as well as a preliminary risk assessment of a prototypal HTR as an input for a comparison of energy supply options. PROTEUS was a critical assembly with an annular driver zone. The central region was filled by arrangements of fuel spheres. The reactivity effect of a water ingress was investigated by simulating the water by polyethylene rods of different diameter inserted into the gaps of a regular package. For sub-criticality measurements in pebble beds, a built-in pulsed neutron source was used. The experimental results were used to validate diffusion and higher order neutron transport models. Concerning thermal hydraulics of gas flows, the vast experience of PSI is focused on hydrogen transport, accumulation, and dispersion in containments of light water reactors. The phenomena are comparable in many aspects to the fluid dynamic issues relevant to HTR. Experiments on hydrogen flows are performed for numerous scenarios in the large-scale containment test facility PANDA. Hydrogen is substituted by helium as a model fluid. An important generic aspect is turbulent mixing in the presence of strong stratification, which is relevant for HTR as well. In a parallel project, generic small-scale mixing experiments with a high density ratio of 1:7 are carried out in a horizontal rectangular channel, where helium and nitrogen flows are brought into contact downstream of the rear edge of a splitter plate. Due to the high density ratio, turbulent mixing is affected by strong non-Boussinesq effects. The measurements taken by Particle Imaging Velocimetry (PIV) and Laser Induced Fluorescence techniques are compared to RANS and LES simulations. Similar large

  14. Development of high temperature gas cooled reactor in China

    Guo, Wentao [Paul Scherrer Institute, Villigen (Switzerland). Dept. of Nuclear Energy and Safety; Schorer, Michael [Swiss Nuclear Forum, Olten (Switzerland)

    2018-02-15

    High temperature gas cooled reactor (HTGR) is one of the six Generation IV reactor types put forward by Generation IV International Forum (GIF) in 2002. This type of reactor has high outlet temperature. It uses Helium as coolant and graphite as moderator. Pebble fuel and ceramic reactor core are adopted. Inherit safety, good economy, high generating efficiency are the advantages of HTGR. According to the comprehensive evaluation from the international nuclear community, HTGR has already been given the priority to the research and development for commercial use. A demonstration project of the High Temperature Reactor-Pebble-�bed Modules (HTR-PM) in Shidao Bay nuclear power plant in China is under construction. In this paper, the development history of HTGR in China and the current situation of HTR-PM will be introduced. The experiences from China may be taken as a reference by the international nuclear community.

  15. Pebbles, Cobbles, and Sockets

    1997-01-01

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

  16. Thermal-hydraulics of helium cooled First Wall channels and scoping investigations on performance improvement by application of ribs and mixing devices

    Arbeiter, Frederik, E-mail: frederik.arbeiter@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, Christian [EUROfusion – Programme Management Unit, Garching (Germany); Chen, Yuming; Ilić, Milica; Schwab, Florian [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Sieglin, Bernhard [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Wenninger, Ronald [EUROfusion – Programme Management Unit, Garching (Germany)

    2016-11-01

    Highlights: • Existing first wall designs and expected plasma heat loads are reviewed. • Heat transfer enhancement methods are investigated by CFD. • The results for heat transfer and friction are given, compared and explained. • Relations for needed pumping power and gained thermal heat are shown. • A range for the maximum permissible heat loads from the plasma is estimated. - Abstract: The first wall (FW) of DEMO is a component with high thermal loads. The cooling of the FW has to comply with the material's upper and lower temperature limits and requirements from stress assessment, like low temperature gradients. Also, the cooling has to be integrated into the balance-of-plant, in a sense to deliver exergy to the power cycle and require a limited pumping power for coolant circulation. This paper deals with the basics of FW cooling and proposes optimization approaches. The effectiveness of several heat transfer enhancement techniques is investigated for the use in helium cooled FW designs for DEMO. Among these are wall-mounted ribs, large scale mixing devices and modified hydraulic diameter. Their performance is assessed by computational fluid dynamics (CFD), and heat transfer coefficients and pressure drop are compared. Based on the results, an extrapolation to high heat fluxes is tried to estimate the higher limits of cooling capabilities.

  17. Creep and fatigue properties of Incoloy 800H in a high-temperature gas-cooled reactor (HTGR) helium environment

    Chow, J.G.Y.; Soo, P.; Epel, L.

    1978-01-01

    A mechanical test program to assess the effects of a simulated HTGR helium environment on the fatigue and creep properties of Incoloy 800H and other primary-circuit metals is described. The emphasis and the objectives of this work are directed toward obtaining information to assess the integrity and safety of an HTGR throughout its service life. The helium test environment selected for study contained 40 μ atm H 2 O, 200 μ atm H 2 , 40 μ atm CO, 10 μ atm CO 2 , and 20 μ atm CH 4 . It is believed that this ''wet'' environment simulates that which could exist in a steam-cycle HTGR containing some leaking steam-generator tubes. A recirculating helium loop operating at about 4 psi in which impurities can be maintained at a constant level, has been constructed to supply the desired environment for fatigue and creep testing

  18. RAMI analysis for DEMO HCPB blanket concept cooling system

    Dongiovanni, Danilo N., E-mail: danilo.dongiovanni@enea.it [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Pinna, Tonio [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Carloni, Dario [KIT, Institute of Neutron Physics and Reactor Technology (INR) – KIT (Germany)

    2015-10-15

    Highlights: • RAMI (reliability, availability, maintainability and inspectability) preliminary assessment for HCPB blanket concept cooling system. • Reliability block diagram (RBD) modeling and analysis for HCPB primary heat transfer system (PHTS), coolant purification system (CPS), pressure control system (PCS), and secondary cooling system. • Sensitivity analysis on system availability performance. • Failure models and repair models estimated on the base of data from the ENEA fusion component failure rate database (FCFRDB). - Abstract: A preliminary RAMI (reliability, availability, maintainability and inspectability) assessment for the HCPB (helium cooled pebble bed) blanket cooling system based on currently available design for DEMO fusion power plant is presented. The following sub-systems were considered in the analysis: blanket modules, primary cooling loop including pipework and steam generators lines, pressure control system (PCS), coolant purification system (CPS) and secondary cooling system. For PCS and CPS systems an extrapolation from ITER Test Blanket Module corresponding systems was used as reference design in the analysis. Helium cooled pebble bed (HCPB) system reliability block diagrams (RBD) models were implemented taking into account: system reliability-wise configuration, operating schedule currently foreseen for DEMO, maintenance schedule and plant evolution schedule as well as failure and corrective maintenance models. A simulation of plant activity was then performed on implemented RBDs to estimate plant availability performance on a mission time of 30 calendar years. The resulting availability performance was finally compared to availability goals previously proposed for DEMO plant by a panel of experts. The study suggests that inherent availability goals proposed for DEMO PHTS system and Tokamak auxiliaries are potentially achievable for the primary loop of the HCPB concept cooling system, but not for the secondary loop. A

  19. Operating windows of pebble divertor

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

    2001-01-01

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

  20. Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

    Cui, Shijie; Zhang, Dalin; Cheng, Jie; Tian, Wenxi; Su, G.H.

    2017-01-01

    As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

  1. Preliminary accident analysis of Loss of Off-Site Power and In-Box LOCA for the CFETR helium cooled solid breeder blanket

    Lian, Qiang; Cui, Shijie [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Zhang, Jing; Zhang, Dalin; Su, G.H. [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China)

    2017-05-15

    Highlights: • The CFETR HCSB blanket has been investigated using RELAP5. • Loss of Off-Site Power is investigated. • The parametric analyses during In-Box LOCA are investigated. • The HCSB blanket for CFETR is designed with sufficient decay heat removal capability. - Abstract: As one of three candidate tritium breeding blanket concepts for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of helium cooled solid breeder (HCSB) blanket was recently proposed. In this paper, the preliminary thermal-hydraulic and safety analyses of the typical outboard equatorial blanket module (No.12) have been carried out using RELAP5/Mod3.4 code. Two design basis accidents are investigated based on the steady-state initialization, including Loss of Off-Site Power and In-Box Loss of Coolant Accident (LOCA). The differences between circulator coast down and circulator rotor locked under Loss of Off-Site Power are compared. Regarding the In-Box LOCA, the influences of different break sizes and locations are thoroughly analyzed based on a relatively accurate modeling method of the heat structures in sub-modules. The analysis results show that the blanket and the combined helium cooling system (HCS) are designed with sufficient decay heat removal capability for both accidents, which can preliminarily verify the feasibility of the conceptual design. The research work can also provide an important reference for parameter optimization of the blanket and its HCS in the next stage.

  2. Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

    Cui, Shijie; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Cheng, Jie; Tian, Wenxi; Su, G.H.

    2017-01-15

    As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

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

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

    2012-12-15

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

  4. Influence of gas pressure on the effective thermal conductivity of ceramic breeder pebble beds

    Dai, Weijing [School of Civil Engineering, The University of Sydney, Sydney (Australia); Pupeschi, Simone [Institute for Applied Materials, Karlsruhe Institute of Technology (KIT) (Germany); Hanaor, Dorian [School of Civil Engineering, The University of Sydney, Sydney (Australia); Institute for Materials Science and Technologies, Technical University of Berlin (Germany); Gan, Yixiang, E-mail: yixiang.gan@sydney.edu.au [School of Civil Engineering, The University of Sydney, Sydney (Australia)

    2017-05-15

    Highlights: • This study explicitly demonstrates the influence of the gas pressure on the effective thermal conductivity of pebble beds. • The gas pressure influence is shown to correlated to the pebble size. • The effective thermal conductivity is linked to thermal-mechanical properties of pebbles and packing structure. - Abstract: Lithium ceramics have been considered as tritium breeder materials in many proposed designs of fusion breeding blankets. Heat generated in breeder pebble beds due to nuclear breeding reaction must be removed by means of actively cooled plates while generated tritiums is recovered by purge gas slowly flowing through beds. Therefore, the effective thermal conductivity of pebble beds that is one of the governing parameters determining heat transport phenomenon needs to be addressed with respect to mechanical status of beds and purge gas pressure. In this study, a numerical framework combining finite element simulation and a semi-empirical correlation of gas gap conduction is proposed to predict the effective thermal conductivity. The purge gas pressure is found to vary the effective thermal conductivity, in particular with the presence of various sized gaps in pebble beds. Random packing of pebble beds is taken into account by an approximated correlation considering the packing factor and coordination number of pebble beds. The model prediction is compared with experimental observation from different sources showing a quantitative agreement with the measurement.

  5. Influence of gas pressure on the effective thermal conductivity of ceramic breeder pebble beds

    Dai, Weijing; Pupeschi, Simone; Hanaor, Dorian; Gan, Yixiang

    2017-01-01

    Highlights: • This study explicitly demonstrates the influence of the gas pressure on the effective thermal conductivity of pebble beds. • The gas pressure influence is shown to correlated to the pebble size. • The effective thermal conductivity is linked to thermal-mechanical properties of pebbles and packing structure. - Abstract: Lithium ceramics have been considered as tritium breeder materials in many proposed designs of fusion breeding blankets. Heat generated in breeder pebble beds due to nuclear breeding reaction must be removed by means of actively cooled plates while generated tritiums is recovered by purge gas slowly flowing through beds. Therefore, the effective thermal conductivity of pebble beds that is one of the governing parameters determining heat transport phenomenon needs to be addressed with respect to mechanical status of beds and purge gas pressure. In this study, a numerical framework combining finite element simulation and a semi-empirical correlation of gas gap conduction is proposed to predict the effective thermal conductivity. The purge gas pressure is found to vary the effective thermal conductivity, in particular with the presence of various sized gaps in pebble beds. Random packing of pebble beds is taken into account by an approximated correlation considering the packing factor and coordination number of pebble beds. The model prediction is compared with experimental observation from different sources showing a quantitative agreement with the measurement.

  6. Research on dynamics and experiments about auxiliary bearings for the helium circulator of the 10 MW high temperature gas-cooled reactor

    Zhao, Yulan; Yang, Guojun; Liu, Xingnan; Shi, Zhengang; Zhao, Lei

    2016-01-01

    Highlights: • The research in this paper is based on the AMB helium circulator of HTR-10. • The dynamic rotor performance is analyzed by processing experimental data. • The mechanical bearing without lubrication can be applied in the HTR-10 system. - Abstract: The 10 MW high-temperature gas-cooled reactor (HTR-10) was constructed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University. The auxiliary bearing is utilized in this system to meet particular requirements for the reactor. The main role of the auxiliary bearing is to constrain rotor displacements and also to support the rotor when the rotor drops down, which is caused by the active magnetic bearing (AMB) failure. The auxiliary bearing needs to endure huge impact, rapid angular acceleration and thermal shock. On the one hand, complex geometrical constructions and forces applied on the system bring difficulties and restrictions to establish an appropriate model to reveal the actual dynamic process. On the other hand, large volumes of data obtained from experiments show velocities and displacements of the rotor during the rotor drop process and then can indicate the actual dynamic interactions to a great extent. The research in this paper is based on the test rig of the AMB helium circulator of HTR-10. This paper aims to analyze the dynamic performance and contact forces of the rotor by processing experimental data. A measurement to estimate forces developed due to impacts of the rotor and the auxiliary bearings is presented. It is of great significance and provides certain foundation to elaborate the rotor drop process for the AMB helium circulator of HTR-10.

  7. An exergoeconomic investigation of waste heat recovery from the Gas Turbine-Modular Helium Reactor (GT-MHR) employing an ammonia–water power/cooling cycle

    Zare, V.; Mahmoudi, S.M.S.; Yari, M.

    2013-01-01

    A detailed exergoeconomic analysis is performed for a combined cycle in which the waste heat from the Gas Turbine-Modular Helium Reactor (GT-MHR) is recovered by an ammonia–water power/cooling cogeneration system. Parametric investigations are conducted to evaluate the effects of decision variables on the performances of the GT-MHR and combined cycles. The performances of these cycles are then optimized from the viewpoints of first law, second law and exergoeconomics. It is found that, combining the GT-MHR with ammonia–water cycle not only enhances the first and second law efficiencies of the GT-MHR, but also it improves the cycle performance from the exergoeconomic perspective. The results show that, when the optimization is based on the exergoeconomics, the unit cost of products is reduced by 5.4% in combining the two mentioned cycles. This is achieved with a just about 1% increase in total investment cost rate since the helium mass flow in the combined cycle is lower than that in the GT-MHR alone. - Highlights: • Application of exergetic cost theory to the combined GT-MHR/ammonia–water cycle. • Enhanced exergoeconomic performance for the combined cycle compared to the GT-MHR. • Comparable investment costs for the combined cycle and the GT-MHR alone

  8. Failure study of helium-cooled tungsten divertor plasma-facing units tested at DEMO relevant steady-state heat loads

    Ritz, G; Pintsuk, G; Linke, J; Hirai, T; Norajitra, P; Reiser, J; Giniyatulin, R; Makhankov, A; Mazul, I

    2009-01-01

    Tungsten was selected as armor material for the helium-cooled divertor in future DEMO-type fusion reactors and fusion power plants. After realizing the design and testing of them under cyclic thermal loads of up to ∼14 MW m -2 , the tungsten divertor plasma-facing units were examined by metallography; they revealed failures such as cracks at the thermal loaded and as-machined surfaces, as well as degradation of the brazing layers. Furthermore, in order to optimize the machining processes, the quality of tungsten surfaces prepared by turning, milling and using a diamond cutting wheel were examined. This paper presents a metallographic examination of the tungsten plasma-facing units as well as technical studies and the characterization on machining of tungsten and alternative brazing joints.

  9. Failure study of helium-cooled tungsten divertor plasma-facing units tested at DEMO relevant steady-state heat loads

    Ritz, G.; Hirai, T.; Norajitra, P.; Reiser, J.; Giniyatulin, R.; Makhankov, A.; Mazul, I.; Pintsuk, G.; Linke, J.

    2009-12-01

    Tungsten was selected as armor material for the helium-cooled divertor in future DEMO-type fusion reactors and fusion power plants. After realizing the design and testing of them under cyclic thermal loads of up to ~14 MW m-2, the tungsten divertor plasma-facing units were examined by metallography; they revealed failures such as cracks at the thermal loaded and as-machined surfaces, as well as degradation of the brazing layers. Furthermore, in order to optimize the machining processes, the quality of tungsten surfaces prepared by turning, milling and using a diamond cutting wheel were examined. This paper presents a metallographic examination of the tungsten plasma-facing units as well as technical studies and the characterization on machining of tungsten and alternative brazing joints.

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

    Setiadipura, T; Irwanto, D; Zuhair, Zuhair

    2015-01-01

    The pebble bed type High Temperature Gas-cooled Reactor (HTGR) is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR) which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO) cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor ...

  11. Thermal analysis of a helium-cooled, tube-bank blanket module for a tandem mirror fusion reactor

    Werner, R.W.

    1983-01-01

    A blanket module concept for the central cell of a tandem mirror reactor is described which takes advantage of the excellent heat transfer and low pressure drop characteristics of tube banks in cross-flow. The blanket employs solid Li 2 O as the tritium breeding material and helium as the coolant. The lithium oxide is contained in tubes arranged within the submodules as a two-pass, cross-flow heat exchanger. Primarily, the heat transfer and thermal-hydraulic aspects of the blanket design study are described in this paper. In particular, the analytical model used for selection of the best tube-bank design parameters is discussed in some detail

  12. Thermal analysis of a helium-cooled, tube-bank blanket module for a tandem-mirror fusion reactor

    Werner, R.W.; Hoffman, M.A.; Johnson, G.L.

    1983-01-01

    A blanket module concept for the central cell of a tandem mirror reactor is described which takes advantage of the excellent heat transfer and low pressure drop characteristics of tube banks in cross-flow. The blanket employs solid Li 2 O as the tritium breeding material and helium as the coolant. The lithium oxide is contained in tubes arranged within the submodules as a two-pass, cross-flow heat exchanger. Primarily, the heat transfer and thermal-hydraulic aspects of the blanket design study are described in this paper. In particular, the analytical model used for selection of the best tube-bank design parameters is discussed in some detail

  13. Parametric Investigation of Brayton Cycle for High Temperature Gas-Cooled Reactor

    Chang Oh

    2004-01-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. In this project, we are investigating helium Brayton cycles for the secondary side of an indirect energy conversion system. Ultimately we will investigate the improvement of the Brayton cycle using other fluids, such as supercritical carbon dioxide. Prior to the cycle improvement study, we established a number of baseline cases for the helium indirect Brayton cycle. These cases look at both single-shaft and multiple-shaft turbomachinery. The baseline cases are based on a 250 MW thermal pebble bed HTGR. The results from this study are applicable to other reactor concepts such as a very high temperature gas-cooled reactor (VHTR), fast gas-cooled reactor (FGR), supercritical water reactor (SWR), and others. In this study, we are using the HYSYS computer code for optimization of the helium Brayton cycle. Besides the HYSYS process optimization, we performed parametric study to see the effect of important parameters on the cycle efficiency. For these parametric calculations, we use a cycle efficiency model that was developed based on the Visual Basic computer language. As a part of this study we are currently investigated single-shaft vs. multiple shaft arrangement for cycle efficiency and comparison, which will be published in the next paper. The ultimate goal of this study is to use supercritical carbon dioxide for the HTGR power conversion loop in order to improve the cycle efficiency to values great than that of the helium Brayton cycle. This paper includes preliminary calculations of the steady state overall Brayton cycle efficiency based on the pebble bed reactor reference design (helium used as the working fluid) and compares those results with an initial calculation of a CO2 Brayton cycle

  14. The Pebble Bed Modular Reactor: An obituary

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

    2011-05-15

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

  15. First operating experience with the helium cooling supply of the superconducting XFEL linear accelerator; Erste Betriebserfahrungen mit der Heliumkaelteversorgung des supraleitenden XFEL-Linearbeschleunigers

    Bozhko, Y.; Escherich, K.; Jensch, K.; Petersen, B.; Schnautz, T.; Sellmann, D. [Deutsches Elektronen-Synchrotron (DESY), Hamburg (Germany); Decker, L.; Ueresin, C.; Zajac, J. [Linde Kryotechnik, Pfungen (Switzerland); Paetzold, T. [Linde Kryotechnik, Hamburg (Germany); ENGIE, Hamburg (Germany)

    2017-07-01

    In Hamburg, the European XFEL project was completed. The superconducting XFEL linear accelerator was commissioned in the course of 2017. The linear accelerator supplies electron bundles with an energy of up to 17.5 GeV and serves as the source of a free-electron laser (XFEL), which provides light with extreme intensity and brilliance with wavelengths in the X-ray range (0.2-0.05 nm). The active part of the linear accelerator consists of 96 cryomodules, each with 8 high-frequency resonators (cavities) and a superconducting magnet packet. The approx. 800 cavities made of high-purity niobium are operated at 1.3 GHz and cooled in a helium II bath at a temperature of 2.0 K. The cavities are surrounded by two thermal shields at temperatures of 5-8 K and 40-80 K. Parallel to the main accelerator, an injector is supplied with two cryogenic modules. The cryogenics of the XFEL linear accelerator includes a helium refrigeration system with design capacities of 2 KW at 2 K, 4 KW at 5/8 K and 24 KW at 40/80 K, various helium transfer lines, a branched distribution system and connection boxes between the module chains in the accelerator tunnel. It is reported on the commissioning of the components and first operating experience. [German] In Hamburg wurde das Europaeische XFEL Projekt fertiggestellt. Der supraleitende XFEL Linearbeschleuniger wurde im Laufe des Jahres 2017 in Betrieb genommen. Der Linearbeschleuniger liefert Elektronenpakete mit einer Energie von bis zu 17,5 GeV und dient als Quelle eines Freien-Elektronen-Lasers (XFEL), der Licht mit extremer Intensitaet und Brillanz mit Wellenlaengen im Roentgenbereich (0,2-0,05 nm) bereitstellt. Der aktive Teil des Linearbeschleunigers besteht aus 96 Kryomodulen mit jeweils 8 Hochfrequenzresonatoren (Kavitaeten) und einem supraleitenden Magnetpaket. Die ca. 800 Kavitaeten aus hochreinem Niob werden bei 1,3 GHz betrieben und in einem Helium-II-Bad bei einer Temperatur von 2,0 K gekuehlt. Die Kavitaeten sind von zwei thermischen

  16. Cool Down Experiences with the SST-1 Helium Cryogenics System before and after Current Feeders System Modification

    Patel, R.; Panchal, P.; Panchal, R.; Tank, J.; Mahesuriya, G.; Sonara, D.; Srikanth, G. L. N.; Garg, A.; Bairagi, N.; Christian, D.; Patel, K.; Shah, P.; Nimavat, H.; Sharma, R.; Patel, J. C.; Gupta, N. C.; Prasad, U.; Sharma, A. N.; Tanna, V. L.; Pradhan, S.

    The SST-1 machine comprises a superconducting magnet system (SCMS), which includes TF and PF magnets. In order to charge the SCMS, we need superconducting current feeders consisting of SC feeders and vapor cooled current leads (VCCLs). We have installed all 10 (+/-) pairs of VCCLs for the TF and PF systems. While conducting initial engineering validation of the SST-1 machine, our prime objective was to produce circular plasma using only the TF system. During the SST-1 campaign I to VI, we have to stop the PF magnets cooling in order to get the cryo- stable conditions for current charging of the TF magnets system. In that case, the cooling of the PF current leads is not essential. It has been also observed that after aborting the PF system cooling, there was a limited experimental window of TF operation. Therefore, in the recent SST-1 campaign-VII, we removed the PF current leads (9 pairs) and kept only single (+/-) pair of the 10,000 A rated VCCLs to realize the charging of the TF system for the extended window of operation. We have observed a better cryogenic stability in the TF magnets after modifications in the CFS. In this paper, we report the comparison of the cool down performance for the SST-1 machine operation before and after modifications of the current feeders system.

  17. A 350 MW HTR with an annular pebble bed core

    Wang Dazhong; Jiang Zhiqiang; Gao Zuying; Xu Yuanhui

    1992-12-01

    A conceptual design of HTR-module with an annular pebble bed core was proposed. This design can increase the unit power capacity of HTR-Module from 200 MWt to 350 MWt while it can keep the inherent safety characteristics of modular reactor. The preliminary safety analysis results for 350 MW HTR are given. In order to solve the problem of uneven helium outlet temperature distribution a gas flow mixing structure at bottom of core was designed. The experiment results of a gas mixing simulation test rig show that the mixing function can satisfy the design requirements

  18. The Emerging Paradigm of Pebble Accretion

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

    2017-01-01

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

  19. Steamworlds: Atmospheric Structure and Critical Mass of Planets Accreting Icy Pebbles

    Chambers, John

    2017-01-01

    In the core accretion model, gas-giant planets first form a solid core, which then accretes gas from a protoplanetary disk when the core exceeds a critical mass. Here, we model the atmosphere of a core that grows by accreting ice-rich pebbles. The ice fraction of pebbles evaporates in warm regions of the atmosphere, saturating it with water vapor. Excess water precipitates to lower altitudes. Beneath an outer radiative region, the atmosphere is convective, following a moist adiabat in saturated regions due to water condensation and precipitation. Atmospheric mass, density, and temperature increase with core mass. For nominal model parameters, planets with core masses (ice + rock) between 0.08 and 0.16 Earth masses have surface temperatures between 273 and 647 K and form an ocean. In more massive planets, water exists as a supercritical convecting fluid mixed with gas from the disk. Typically, the core mass reaches a maximum (the critical mass) as a function of the total mass when the core is 2–5 Earth masses. The critical mass depends in a complicated way on pebble size, mass flux, and dust opacity due to the occasional appearance of multiple core-mass maxima. The core mass for an atmosphere of 50% hydrogen and helium may be a more robust indicator of the onset of gas accretion. This mass is typically 1–3 Earth masses for pebbles that are 50% ice by mass, increasing with opacity and pebble flux and decreasing with pebble ice/rock ratio.

  20. Steamworlds: Atmospheric Structure and Critical Mass of Planets Accreting Icy Pebbles

    Chambers, John, E-mail: jchambers@carnegiescience.edu [Carnegie Institution for Science Department of Terrestrial Magnetism, 5241 Broad Branch Road, NW, Washington, DC 20015 (United States)

    2017-11-01

    In the core accretion model, gas-giant planets first form a solid core, which then accretes gas from a protoplanetary disk when the core exceeds a critical mass. Here, we model the atmosphere of a core that grows by accreting ice-rich pebbles. The ice fraction of pebbles evaporates in warm regions of the atmosphere, saturating it with water vapor. Excess water precipitates to lower altitudes. Beneath an outer radiative region, the atmosphere is convective, following a moist adiabat in saturated regions due to water condensation and precipitation. Atmospheric mass, density, and temperature increase with core mass. For nominal model parameters, planets with core masses (ice + rock) between 0.08 and 0.16 Earth masses have surface temperatures between 273 and 647 K and form an ocean. In more massive planets, water exists as a supercritical convecting fluid mixed with gas from the disk. Typically, the core mass reaches a maximum (the critical mass) as a function of the total mass when the core is 2–5 Earth masses. The critical mass depends in a complicated way on pebble size, mass flux, and dust opacity due to the occasional appearance of multiple core-mass maxima. The core mass for an atmosphere of 50% hydrogen and helium may be a more robust indicator of the onset of gas accretion. This mass is typically 1–3 Earth masses for pebbles that are 50% ice by mass, increasing with opacity and pebble flux and decreasing with pebble ice/rock ratio.

  1. Experimental measurement of effective thermal conductivity of packed lithium-titanate pebble bed

    Mandal, D.; Sathiyamoorthy, D.; Vinjamur, M.

    2012-01-01

    Lithium titanate is a promising solid breeder material for the fusion reactor blanket. Packed lithium titanate pebble bed is considered for the blanket. The thermal energy; that will be produced in the bed during breeding and the radiated heat from the reactor core absorbed must be removed. So, the experimental thermal property data are important for the blanket design. In past, a significant amount of works were conducted to determine the effective thermal conductivity of packed solid breeder pebble bed, in helium atmosphere, but no flow of gas was considered. With increase in gas flow rate, effective thermal conductivity of pebble bed increases. Particle size and void fraction also affect the thermal properties of the bed significantly. An experimental facility with external heat source was designed and installed. Experiments were carried out with lithium-titanate pebbles of different sizes at variable gas flow rates and at different bed wall temperature. It was observed that effective thermal conductivity of pebble bed is a function of particle Reynolds number and temperature. From the experimental data two correlations have been developed to estimate the effective thermal conductivity of packed lithium-titanate pebble bed for different particle Reynolds number and at different temperatures. The experimental details and results are discussed in this paper.

  2. Operating Manual of Helium Refrigerator (Rev. 2)

    Song, K.M.; Son, S.H.; Kim, K.S.; Lee, S.K.; Kim, M.S. [Korea Electric Power Research Institute, Taejon (Korea)

    2002-07-01

    A helium refrigerator was installed as a supplier of 20K cold helium to the cryogenic distillation system of WTRF pilot plant. The operating procedures of the helium refrigerator, helium compressor and auxiliary apparatus are described for the safety and efficient operation in this manual. The function of the helium refrigerator is to remove the impurities from the compressed helium of about 250psig, to cool down the helium from ambient temperature to 20K through the heat exchanger and expansion engine and to transfer the cold helium to the cryogenic distillation system. For the smoothly operation of helium refrigerator, the preparation, the start-up, the cool-down and the shut-down of the helium refrigerator are described in this operating manual. (author). 3 refs., 14 tabs.

  3. Variable electricity and steam from salt, helium and sodium cooled base-load reactors with gas turbines and heat storage - 15115

    Forsberg, C.; McDaniel, P.; Zohuri, B.

    2015-01-01

    Advances in utility natural-gas-fired air-Brayton combed cycle technology is creating the option of coupling salt-, helium-, and sodium-cooled nuclear reactors to Nuclear air-Brayton Combined Cycle (NACC) power systems. NACC may enable a zero-carbon electricity grid and improve nuclear power economics by enabling variable electricity output with base-load nuclear reactor operations. Variable electricity output enables selling more electricity at times of high prices that increases plant revenue. Peak power is achieved using stored heat or auxiliary fuel (natural gas, bio-fuels, hydrogen). A typical NACC cycle includes air compression, heating compressed air using nuclear heat and a heat exchanger, sending air through a turbine to produce electricity, reheating compressed air, sending air through a second turbine, and exhausting to a heat recovery steam generator (HRSG). In the HRSG, warm air produces steam that is used to produce added electricity. For peak power production, auxiliary heat (natural gas, stored heat) is added before the air enters the second turbine to raise air temperatures and power output. Like all combined cycle plants, water cooling requirements are dramatically reduced relative to other power cycles because much of the heat rejection is in the form of hot air. (authors)

  4. The ESKOM pebble bed modular reactor

    Gittus, J.H.

    1999-01-01

    An audit has been made of the design, construction, safety, economics and marketability of the ESKOM pebble bed modular reactor (PBMR). In this paper that audit is briefly summarized. The principal conclusions of the audit are as follows. The design is sound. It is a logical development of the designs proposed for other, modern, high-temperature gas-cooled reactors. More than 80% of the cost of constructing and commissioning a series of PBMRs would be spent in South Africa. The PBMR is much safer than existing nuclear power reactors and for many practical purposes it may be treated as a conventional chemical plant. The PBMR is economically competitive with thermal power stations. There is a substantial global market for the PBMR. (author)

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

    Ion, S.; Nicholls, D.; Matzie, R.; Matzner, D.

    2004-01-01

    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)

  6. Advances in conceptual design of a gas-cooled accelerator driven system (ADS) transmutation devices to sustainable nuclear energy development

    Garcia, Rosales; Fajardo, Garcia; Curbelo, Perez; Oliva, Munoz; Hernandez, Garcia; Castells, Escriva; Abanades

    2011-01-01

    The possibilities of a nuclear energy development are considerably increasing with the world energetic demand increment. However, the management of nuclear waste from conventional nuclear power plants and its inventory minimization are the most important issues that should be addressed. Fast reactors and Accelerator Driven Systems (ADS) are the main options to reduce the long-lived radioactive waste inventory. Pebble Bed Very High Temperature advanced systems have great perspectives to assume the future nuclear energy development challenges. The conceptual design of a Transmutation Advanced Device for Sustainable Energy Applications (TADSEA) has been made in preliminary studies. The TADSEA is an ADS cooled by helium and moderated by graphite that uses as fuel small amounts of transuranic elements in the form of TRISO particles, confined in 3 cm radius graphite pebbles forming a pebble bed configuration. It would be used for nuclear waste transmutation and energy production. In this paper, the results of a method for calculating the number of whole pebbles fitting in a volume according to its size are showed. From these results, the packing fraction influence on the TADSEAs main work parameters is studied. In addition, a redesign of the previous configuration, according to the established conditions in the preliminary design, i.e. the exit thermal power, is made. On the other hand, the heterogeneity of the TRISO particles inside the pebbles can not be negligible. In this paper, a study of the power density distribution inside the pebbles by means of a detailed simulation of the TRISO fuel particles and using an homogeneous composition of the fuel is addressed. (author)

  7. Analysis of impact of mixing flow on the pebble bed high temperature reactor

    Hao Chen; Li Fu; Guo Jiong

    2014-01-01

    The impact of the mixing flow in the pebble flow on pebble bed high temperature gas cooled reactor (HTR) was analyzed in the paper. New code package MFVSOP which can simulate the mixing flow was developed. The equilibrium core of HTR-PM was selected as reference case, the impact of the mixing flow on the core parameters such as core power peak factor, power distribution was analyzed with different degree of mixing flow, and uncertainty analysis was carried out. Numerical results showed that the mixing flow had little impact on key parameters of pebble bed HTR, and the multiple-pass-operation-mode in pebble bed HTR can reduce the uncertainty arouse from the mixing flow. (authors)

  8. Neutronic performance of two European breeder-inside-tube (BIT) blankets for DEMO: the helium-cooled ceramic LiAlO2 with Be multiplier and the water-cooled liquid Li17Pb

    Petrizzi, L.; Rado, V.

    1995-01-01

    In support of ENEA activity in the European Community Test Programme, neutron analysis has been performed on the two latest blanket designs: helium-cooled ceramic breeder-inside-tube (BIT) (with LiAlO 2 and Be multiplier) and water-cooled liquid Li 17 Pb in cylindrical modules (CM). The powerful MCNP Monte Carlo code was used (version 4.2). A detailed and accurate description of the geometrical model has been performed by inserting the main reactor details and avoiding breeder material dilution inside the modules. The tritium breeding ratio (TBR) performance is low for the solid breeder BIT blanket (with 10 ports 1.011) due mainly to low blanket coverage near the exhaust duct, and this solution should be revised. The CM Li 17 Pb blanket reaches a sufficient TBR (1.059, with ports) to rely on tritium self-sufficiency. Shielding properties, with respect to the toroidal field coils, have been estimated in a simplified model by means of the ANISN code, supplied with a nuclear data library consistent with that used by MCNP. The analysis suggests that a careful shield thickness/composition design should be used to ensure the shielding capability of the whole blanket plus shield system. (orig.)

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

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

    2010-01-01

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

  10. Dynamics of a small direct cycle pebble bed HTR

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

    2001-01-01

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

  11. Thermodynamic properties of helium in the range from 20 to 15000C and 1 to 100 bar. Reactor core design of high-temperature gas-cooled reactors. Pt. 1

    Kipke, H.E.; Stoehr, A.; Banerjea, A.; Hammeke, K.; Huepping, N.

    1978-12-01

    The following report presents in tabular form the safety standard of the nuclear safety standard commission (KTA) on reactor core design of high-temperature gas-cooled reactors. Part 1: Calculation of thermodynamic properties of helium The basis of the present work is the data and formulae given by H. Petersen for the calculation of density, specific heat, thermal conductivity and dynamic viscosity of helium together with the formula for their standard deviations in the range of temperature and pressure stated above. The relations for specific enthalpy and specific entropy have been derived from density and specific heat, whereby specific heat is assumed constant over the given range of temperature and pressure. The latter section of this report contains tables of thermodynamic properties of helium calculated from the equations stated earlier in this paper. (orig.) [de

  12. Numerical modelling for the effective thermal conductivity of lithium meta titanate pebble bed with different packing structures

    Panchal, Maulik, E-mail: maulikpanchal@ipr.res.in [Institute for Plasma Research, Bhat, Gandhinagar-382428 (India); Chaudhuri, Paritosh [Institute for Plasma Research, Bhat, Gandhinagar-382428 (India); Van Lew, Jon T; Ying, Alice [UCLA, MAE Department, Los Angeles, CA 90095-1597 (United States)

    2016-11-15

    Highlights: • The effective thermal conductivity (k{sub eff}) of lithium meta-titanate (Li{sub 2}TiO{sub 3}) pebble beds is an important parameter for the design and analysis of TBM in ITER. • The k{sub eff} of Li{sub 2}TiO{sub 3} pebble beds under stagnant helium gas have been determined numerically using different uniform packing structures and random close packing (RCP) structures. • k{sub eff} of Li{sub 2}TiO{sub 3} pebble beds with different packing fractions have been reported as function of temperature; k{sub eff} of the RCP Li{sub 2}TiO{sub 3} pebble bed is compared with reported experimental results. • The numerically-determined k{sub eff} of the RCP Li{sub 2}TiO{sub 3} pebble bed agrees reasonably well with the experimental data and Zehner-Schlunder correlation. - Abstract: The effective thermal conductivity (k{sub eff}) of lithium meta-titanate (Li{sub 2}TiO{sub 3}) pebble beds is an important parameter for the design and analysis of IN LLCB TBM (Indian Lead Lithium Ceramic Breeder Test Blanket Module). The k{sub eff} of Li{sub 2}TiO{sub 3} pebble beds under stagnant helium gas have been determined numerically using different uniform packing structures and random close packing (RCP) structures. The uniform packing structures of Li{sub 2}TiO{sub 3} pebble bed are modelled by using the simple cubic, body centered cubic and face centered cubic arrangement. The packing structure of the RCP bed of Li{sub 2}TiO{sub 3} pebbles is generated with the discrete element method (DEM) code. k{sub eff} of Li{sub 2}TiO{sub 3} pebble beds with different packing fractions have been reported as function of temperature; k{sub eff} of the RCP Li{sub 2}TiO{sub 3} pebble bed is compared with reported experimental results from literature. The numerically determined k{sub eff} of the Li{sub 2}TiO{sub 3} pebble bed agrees reasonably well with the experimental data.

  13. Pebble Bed Reactor Dust Production Model

    Abderrafi M. Ougouag; Joshua J. Cogliati

    2008-09-01

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

  14. Pebble Bed Reactor Dust Production Model

    Abderrafi M. Ougouag; Joshua J. Cogliati

    2008-01-01

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

  15. A method of lines solution of the transient behavior of the helium cooled power leads for the SSC

    Demko, J.A.; Schiesser, W.E.; Carcagno, R.; McAshan, M.

    1995-01-01

    In this study, a detailed numerical thermal mode of a 6.5 kA power lead for the Superconducting Super Collider has been developed, which was adapted from the dynamic model developed by Schiesser. The transient behavior of the power leads was modeled using, a method of lines (MOL) approach. The model was developed to pmvide a tool for analyzing coolant control strategies as well as an understanding of the behavior of the leads under presumed system transients. Results for a current ramp up to 4970 amps are favorably compared with measurements. Also, a loss of cooling situation is predicted to determine the transient temperature distribution under an off-design condition

  16. Pressure transients analysis of a high-temperature gas-cooled reactor with direct helium turbine cycle

    Dang, M.; Dupont, J. F.; Jacquemoud, P.; Mylonas, R. [Eidgenoessisches Inst. fuer Reaktorforschung, Wuerenlingen (Switzerland)

    1981-01-15

    The direct coupling of a gas cooled reactor with a closed gas turbine cycle leads to a specific dynamic plant behaviour, which may be summarized as follows: a) any operational transient involving a variation of the core mass flow rate causes a variation of the pressure ratio of the turbomachines and leads unavoidably to pressure and temperature transients in the gas turbine cycle; and b) very severe pressure equalization transients initiated by unlikely events such as the deblading of one or more turbomachines must be taken into account. This behaviour is described and illustrated through results gained from computer analyses performed at the Swiss Federal Institute for Reactor Research (EIR) in Wurenlingen within the scope of the Swiss-German HHT project.

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

    1978-09-01

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

  18. Thermal-hydraulic performance of a multiple jet cooling module with a concave dimple array in a helium-cooled divertor

    Jung, Hyo-Yeon; Kim, Kwang-Yong, E-mail: kykim@inha.ac.kr

    2017-01-15

    A numerical study was performed to evaluate the thermal-hydraulic performance of a finger type cooling module, where multiple jets impinge on the surface with concave dimples, in the divertor of a nuclear fusion reactor. Conjugate heat transfer was analyzed in both the solid and fluid domains using three-dimensional Reynolds-averaged Navier-Stokes equations with the shear stress transport turbulence model. The computational domain consisted of a single fluid domain and three solid domains: tile, thimble, and cartridge. The numerical results for the temperature variation on the tile were validated in comparison with the experimental data. A parametric study was performed with two design variables, the ratios of dimple diameter and dimple height to the nozzle diameter, and two dimple arrays (inline and staggered arrays). The parametric study showed that the heat transfer rate was increased by up to 2.62% by introducing concave dimples, and that the heat transfer and pressure drop performances increased with increasing diameter and height of the dimples for a specified dimple array.

  19. Radiation Protection Practices during the Helium Circulator Maintenance of the 10 MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10

    Chengxiang Guo

    2016-01-01

    Full Text Available Current radiation protection methodology offers abundant experiences on light-water reactors, but very few studies on high temperature gas-cooled reactor (HTR. To fill this gap, a comprehensive investigation was performed to the radiation protection practices in the helium circulator maintenance of the Chinese 10 MW HTR test module (HTR-10 in this paper. The investigation reveals the unique behaviour of HTR-10’s radiation sources in the maintenance as well as its radionuclide species and presents the radiation protection methods that were tailored to these features. Owing to these practices, the radioactivity level was kept low throughout the maintenance and only low-level radioactive waste was generated. The quantitative analysis further demonstrates that the decontamination efficiency was over 89% for surface contamination and over 34% for γ dose rate and the occupational exposure was much lower than both the limits of regulatory and the exposure levels in comparable literature. These results demonstrate the effectiveness of the reported radiation protection practices, which directly provides hands-on experience for the future HTR-PM reactor and adds to the completeness of the radiation protection methodology.

  20. 3D Nondestructive Visualization and Evaluation of TRISO Particles Distribution in HTGR Fuel Pebbles Using Cone-Beam Computed Tomography

    Gongyi Yu

    2017-01-01

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

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

    Irwanto, Dwi; Setiadipura, Topan; Pramutadi, Asril

    2017-07-01

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

  2. A comparative study on the effective thermal conductivity of a single size beryllium pebble bed

    Abou-Sena, A.; Ying, A.; Abdou, M.

    2004-01-01

    Solid breeder blankets generally use beryllium-helium pebble beds to ensure sufficient tritium breeding. The data of the effective thermal conductivity, k eff , of beryllium pebble beds is important to the design of fusion blankets. It serves as a database for benchmarking the models of pebble beds. The objective of this paper is to review and compare the available data (obtained by several studies) of the effective thermal conductivity of beryllium pebble beds in order to address the current status of these data. Two comparisons are presented: one for the data of k eff versus bed mean temperature and the second one for the data of k eff versus external applied pressures. The data (k eff versus bed temperature) reported by Enoeda et al., Dalle Donne et al., and UCLA, have a similar particle size and packing fraction. Despite their similarity, the standard deviation values of their data are around 32%. Also, the data of the effective thermal conductivity as a function of mechanical pressure have standard deviation values of ∼50%. From the presented comparisons, significant discrepancies among the available data of k eff of the beryllium pebble beds were observed. These discrepancies may be attributed to the apparent differences among available studies, such as experiment technique, packing fraction, particle characteristics, bed dimensions, and temperature range and gradient across the bed. (author)

  3. The R&D of HTGR high temperature helium sampling loop: From HTR-10 to HTR-PM

    Fang, Chao, E-mail: fangchao@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Collaborative Innovation Center of Advanced Nuclear Energy Technology, Tsinghua University, Beijing 100084 (China); The Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing 100084 (China); Bao, Xuyin; Yang, Chen; Yang, Yanran; Cao, Jianzhu [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Collaborative Innovation Center of Advanced Nuclear Energy Technology, Tsinghua University, Beijing 100084 (China); The Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing 100084 (China)

    2016-09-15

    A High Temperature Helium Sampling Loop (HTHSL) for studying the transportation (deposition) behavior and total amount of solid fission products in high-temperature helium coming from the steam generator (SG) in the 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) and High Temperature Reactor-Pebble bed Modules (HTR-PM) are researched and designed, respectively. Through the optimal design and simulation based on thermohydraulics analysis, the three-sleeve structure of deposition sampling device (DSD) could realize full-length temperature control evenly so that it could be used to study fission products in the primary circuit of HTR-10. On the other hand, an improved DSD is also designed for HTR-PM based on corresponding simulations, which could be used to sample the important nuclei in the high temperature helium from SG. These schemes offer two different methods to obtain the original source term in the high temperature helium, which will provide deeper understanding for the analysis of source terms of HTGR.

  4. Analysis of multi-scale spatial separation in a block-type thorium-loaded helium-cooled high-temperature reactor

    Huang, Jie; Ding, Ming

    2017-01-01

    Highlights: • Four-level of spatial separation is described in a block-type thorium-loaded HTR. • A traditional two-step calculation scheme is used to get the neutronic performance. • Fuel cycle cost is calculated by the levelised lifetime cost method. • Fuel cycle cost decreases with the increase of separation level or thorium content. • Effective enrichment basically determines the fuel cycle cost. - Abstract: With nuclear energy’s rapid development in recent years, supply of nuclear fuel has become increasingly important. Thorium has re-gained attention because of its abundant reserves and excellent physical properties. Compared to the homogeneous Th/U MOX fuel, separation of thorium and uranium in space is a better use of thorium. Therefore, this paper describes four-level spatial separation – no separation, tristructural-isotropic (TRISO) level, channel level and block level – in a block-type thorium-loaded helium-cooled high-temperature reactor (HTR). A traditional two-step calculation scheme, lattice calculation followed by core calculation, is used to get the neutronic performance of the equilibrium cycle, including uranium enrichment, mass of fuel, effective multiplication factor, and average conversion ratio. Based on these data, the fuel cycle cost of different-scale spatial separation can be calculated by the levelised lifetime cost method as a function of thorium content. As the separation level increases from no separation to channel level, the effective enrichment decreases 15% due to the increase of resonance escape probability. So there is a 13% drop for the fuel cycle cost. For TRISO-level separation, as the thorium content increases from 9 to 57%, the effective enrichment decreases 14% because of the superior breeding capacity of U-233. As a result, the fuel cycle cost also has about a 12% decrease. From the perspective of fuel cycle economics, channel-level separation with 60% thorium content is suggested.

  5. The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant

    Ariani, Menik, E-mail: menikariani@gmail.com; Satya, Octavianus Cakra; Monado, Fiber [Department of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University, jl Palembang-Prabumulih km 32 Indralaya OganIlir, South of Sumatera (Indonesia); Su’ud, Zaki [Nuclear and Biophysics Research Division, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, jlGanesha 10, Bandung (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, 2-12-11N1-17 Ookayama, Meguro-Ku, Tokyo (Japan)

    2016-03-11

    The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactors with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on “Region-8” and “Region-10” core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years).

  6. A solid target for SINQ based on a Pb-shot Pebble-bed

    Atchison, F.; Heidenreich, G.

    1991-01-01

    Preliminary results from scoping calculations examining the possibilities of implementing a Pebble-bed of Pb-shot as a target for SINQ are presented. The primary design objects are set out and estimates of heating and activation given. Cooling circuit parameters are discussed and estimates for operating conditions presented. A short discussion of problems associated with a realisation is included. (author)

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

    Wei Chunlin; Zhang Jian; Shan Wenzhi; Jing Xingqing

    2013-01-01

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

  8. Post irradiation characterization of beryllium and beryllides after high temperature irradiation up to 3000 appm helium production in HIDOBE-01

    Fedorov, A.V., E-mail: fedorov@nrg.eu [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, Petten, 1755 ZG (Netherlands); Til, S. van; Stijkel, M.P. [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, Petten, 1755 ZG (Netherlands); Nakamichi, M. [Japan Atomic Energy Agency, Rokkasho (Japan); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/ Josep Pla, n° 2, Torres Diagonal Litoral, Edificio B3, Barcelona 08019 (Spain)

    2016-01-15

    Titanium beryllides are considered as advanced candidate material for neutron multiplier for the helium cooled pebble bed (HCPB) and/or the water cooled ceramic breeder (WCCB) breeder blankets. In the HIDOBE-01 (HIgh DOse irradiation of BEryllium) experiment, beryllium and beryllide pellets with 5 at% and 7 at% Ti are irradiated at four different target temperatures (T{sub irr}): 425 °C, 525 °C, 650 °C and 750 °C up to the dose corresponding to 3000 appm He production in beryllium. The pellets were supplied by JAEA. During post irradiation examinations the critical properties of volumetric swelling and tritium retention were studied. Both titanium beryllide grades show significantly less swelling than the beryllium grade, with the difference increasing with the irradiation temperature. The irradiation induced swelling was studied by using direct dimensions. Both beryllide grades showed much less swelling as compare to the reference beryllium grade. Densities of the grades were studied by Archimedean immersion and by He-pycnometry, giving indications of porosity formation. While both beryllide grades show no significant reduction in density at all irradiation temperatures, the beryllium density falls steeply at higher T{sub irr}. Finally, the tritium release and retention were studied by temperature programmed desorption (TPD). Beryllium shows the same strong tritium retention as earlier observed in studies on beryllium pebbles, while the tritium inventory of the beryllides is significantly less, already at the lowest T{sub irr} of 425 °C.

  9. Research and application of packing density for pebble bed in HTR

    Yu Fujiang; Xie Fei; Sun Ximing

    2015-01-01

    The pebble bed high temperature gas-cooled reactor is one of the major types of reactors developed by Chinese nuclear technology. The statistical analysis for packing density in the pebble bed is an important issue of physical-thermal calculation and safety analysis. Aimed to this problem, a new kind of method was set up to solve this problem. Compared with the traditional lattice-fill method and the experiment, its efficiency and accuracy were verified, while helping to find out the best length of unit in the traditional lattice-fill method. This method was used to analyze the boundary effects observed by experiments. (authors)

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

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

    2013-07-01

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

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

    Li, Y.; Ji, W.

    2013-01-01

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

  12. Liquid helium

    Atkins, K R

    1959-01-01

    Originally published in 1959 as part of the Cambridge Monographs on Physics series, this book addresses liquid helium from the dual perspectives of statistical mechanics and hydrodynamics. Atkins looks at both Helium Three and Helium Four, as well as the properties of a combination of the two isotopes. This book will be of value to anyone with an interest in the history of science and the study of one of the universe's most fundamental elements.

  13. A solution for the helium problem. Cryogen-free cooling systems for low temperatures; Eine Loesung fuer das Heliumproblem. Kryogenfreie Kuehlsysteme fuer tiefe Temperaturen

    Good, Jeremy [Cryogenic Limited, London (United Kingdom)

    2014-09-15

    Pulse tube or Gifford-McMahon coolers are related to Stirling engines. Extremely low temperatures - 1 K can be reached with these devices. As a cryogen-free system the devices need only small amounts of helium as working gas. This fact reduces the gaseous and liquid helium consumption of research labs considerably and allows new applications. The cost-efficiency of this alternative technique is important for research facilities that use superconducting magnets.

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

    Meng Xianke; Sun Zhongning; Xu Guangzhan

    2012-01-01

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

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

    Setiadipura, T.; Zuhair; Irwanto, D.

    2015-01-01

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

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

    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

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

    Cogliati, Joshua J.; Ougouag, Abderrafi M.

    2010-01-01

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

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

    Meng Xianke; Sun Zhongning; Zhou Ping; Xu Guangzhan

    2012-01-01

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

  19. Helium Loop for the HCPB Test Blanket Module

    Neuberger, H.; Boccaccini, L.V.; Ghidersa, B. E.; Jin, X.; Meyder, R.

    2006-01-01

    In the frame of the activities of the EU Breeder Blanket Programme and of the Test Blanket Working Group, the Helium loop for the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) in ITER has been investigated with regard to the layout definition, selection of components, control, dimensioning and integration. This paper presents the status of development. The loop design for the HCPB-TBM in ITER will mainly base on the experience gained from Helium Loop Karlsruhe (HELOKA) which is currently developed at the FZK for experiments under ITER relevant conditions. The ITER loop will be equipped with similar components like HELOKA and will mainly consist of a circulator with variable speed drive, a recuperator, an electric heater, a cooler, a dust filter and auxilary components e.g. pipework and valves. A Coolant Purification System (CPS) and a Pressure Control System (PCS) are foreseen to meet the requirements on coolant conditioning. To prepare a TBM for a new experimental campaign, a succession of operational states like '' cold maintenance '', '' baking '' and '' cold standby '' is required. Before a pulse operation, a '' hot stand-by '' state should be achieved providing the TBM with inlet coolant at nominal conditions. This operation modus is continued in the dwell time waiting for the successive pulse. A '' tritium out-gassing '' will be also required after several TBM-campaigns to remove the inventory rest of T in the beds for measurement purpose. The dynamic circuit behaviour during pulses, transition between different operational states as well as the behaviour in accident situations are investigated with RELAP. The main components of the loop will be accommodated inside the Tokamak Cooling Water System(TCWS)- vault from where the pipes require connection to the TBM which is attached to port 16 of the vacuum vessel. Therefore pipes across the ITER- building of about 110 m in length (each) are required. Additional equipment is also located in the port cell

  20. Investigation of effective thermal conductivity for pebble beds by one-way coupled CFD-DEM method for CFETR WCCB

    Chen, Lei [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); University of Science and Technology of China, Hefei, Anhui 230027 (China); Chen, Youhua [University of Science and Technology of China, Hefei, Anhui 230027 (China); Huang, Kai [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); University of Science and Technology of China, Hefei, Anhui 230027 (China)

    2016-05-15

    Highlights: • A CFD-DEM coupled numerical model is built based on the prototypical blanket pebble bed. • The numerical model can be applied to simulate heat transfer of a pebble bed and estimate effective thermal conductivity. • The numerical model agrees well with the theoretical SZB model. • Effective thermal conductivity of pebble beds for WCCB is estimated by the current model. - Abstract: The mono-sized beryllium pebble bed and the multi-sized Li{sub 2}TiO{sub 3}/Be{sub 12}Ti mixed pebble bed are the main schemes for the Water-cooled ceramic breeder blanket (WCCB) of China Fusion Engineering Test Reactor (CFETR). And the effective thermal conductivity (k{sub eff}) of the pebble beds is important to characterize the thermal performance of WCCB. In this study, a one-way coupled CFD-DEM method was employed to simulate heat transfer and estimate k{sub eff}. The geometric topology of a prototypical blanket pebble bed was produced by the discrete element method (DEM). Based on the geometric topology, the temperature distribution and the k{sub eff} were obtained by the computational fluid dynamics (CFD) analysis. The current numerical model presented a good performance to calculate k{sub eff} of the beryllium pebble bed, and according to the modeling of the Li{sub 2}TiO{sub 3}/Be{sub 12}Ti mixed pebble bed, k{sub eff} was estimated with values ranged between 2.0 and 4.0 W/(m∙K).

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

    Pohl, P.

    2006-01-01

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

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

    Xu Yong; Li Yanjie

    2005-01-01

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

  3. Effect of a flow-corrective insert on the flow pattern in a pebble bed reactor

    Li, Yu; Gui, Nan; Yang, Xingtuan [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China); Tu, Jiyuan [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China); School of Aerospace, Mechanical & Manufacturing Engineering, RMIT University, Melbourne 3083, VIC (Australia); Jiang, Shengyao, E-mail: shengyaojiang@sina.com [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2016-04-15

    Highlights: • Effect of an insert on improving flow uniformity and eliminating stagnant zone is studied. • Three values concerned with the stagnant zone, radial uniformity and flow sequence are used. • Outlet diameter is a critical parameter that determines balancing mechanism of the insert. • Height/location is varied to let the insert work in unbalanced region and avoid adverse effect. - Abstract: A flow-corrective insert is adopted in the pebble-bed high temperature gas-cooled reactor (HTGR) to improve flow performance of the pebble flow for the first time. 3D discrete element method (DEM) modeling is employed to study this slow and dense granular flow. It is verified that locating a properly designed insert in the bed can help transform unsatisfactory flow field to the preferred flow pattern for pebble bed reactors. Three characteristic values on the stagnant zone, radial uniformity and flow sequence of pebble flow are defined to evaluate uniformity of the overall flow field quantitatively. The results demonstrate that the pebble bed equipped with an insert performs better than normal beds from all these three aspects. Moreover, based on numerical experiments, several universal tips for insert design on height, location and outlet diameter are suggested.

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

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

    2008-01-01

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

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

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

    2008-01-01

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

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

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

    1997-01-01

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

  7. Helium cryogenics

    Van Sciver, Steven W

    2012-01-01

    Twenty five years have elapsed since the original publication of Helium Cryogenics. During this time, a considerable amount of research and development involving helium fluids has been carried out culminating in several large-scale projects. Furthermore, the field has matured through these efforts so that there is now a broad engineering base to assist the development of future projects. Helium Cryogenics, 2nd edition brings these advances in helium cryogenics together in an updated form. As in the original edition, the author's approach is to survey the field of cryogenics with emphasis on helium fluids. This approach is more specialized and fundamental than that contained in other cryogenics books, which treat the associated range of cryogenic fluids. As a result, the level of treatment is more advanced and assumes a certain knowledge of fundamental engineering and physics principles, including some quantum mechanics. The goal throughout the work is to bridge the gap between the physics and engineering aspe...

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

    Bende, E.E

    2000-01-24

    parameters like the temperature coefficients of reactivity, the delayed neutron fractions and the mean neutron generation time as well as their evolutions during burnup are presented. The calculations for this parameter study still refer to an infinite pebble bed or, more precisely, to a 'pebble-cell' with isotropically reflecting boundaries. On the basis of the results of the cell calculations, two pebble configurations were chosen as viable candidates for fuel in a real reactor. These configurations embrace pebbles with either 1 g or 2 g Pu/pebble contained by coated particles with fuel kernel diameters of 220 {mu}m. Chapter 5 presents a conceptual design of a pebble-bed HTR that is operated with pebbles having the mentioned Pu-loadings. This design, which will be designated as the HTR-Pu, is a modified version of the Dutch ACACIA design, which is a small pebble-bed reactor, based on the Peu-a-Peu fuelling principle, with a thermal power of 40 MWth. The proposed reactor design is only a first try and has not been subject to any optimisation procedures. The calculations are performed by the PANTHERMIX-code system which establishes a coupling between the diffusion code PANTHER and the thermal-hydraulics code THERMIX/DIREKT. Section 5.4 reports the calculational results during normal operation. This embraces burnup calculations that simulate the continuous fuelling of pebbles on top of the initial pebble bed until the maximum core height is reached. Section 5.5 focuses on the excess reactivities, the control rod requirements and an envisaged start-up procedure of the HTR-Pu, for both Pu-loadings. Finally, section 5.6 shows the calculational results of the HTR-Pu under off-normal conditions. Several loss-of-cooling scenarios were simulated in order to examine whether the HTR's passive safety features are preserved in case the reactor is loaded with the mentioned Pu-fuels. Appendix A presents an analytic expression for the average Dancoff factor of a fuel kernel

  9. Changes in the metallicity of gas giant planets due to pebble accretion

    Humphries, R. J.; Nayakshin, S.

    2018-06-01

    We run numerical simulations to study the accretion of gas and dust grains on to gas giant planets embedded into massive protoplanetary discs. The outcome is found to depend on the disc cooling rate, planet mass, grain size, and irradiative feedback from the planet. If radiative cooling is efficient, planets accrete both gas and pebbles rapidly, open a gap, and usually become massive brown dwarfs. In the inefficient cooling case, gas is too hot to accrete on to the planet but pebble accretion continues and the planets migrate inward rapidly. Radiative feedback from the planet tends to suppress gas accretion. Our simulations predict that metal enrichment of planets by dust grain accretion inversely correlates with the final planet mass, in accordance with the observed trend in the inferred bulk composition of Solar system and exosolar giant planets. To account for observations, however, as many as ˜30-50 per cent of the dust mass should be in the form of large grains.

  10. Effects of hydrogen mixture into helium gas on deuterium removal from lithium titanate

    Tsuchiya, Akihito, E-mail: tsuchiya@frontier.hokudai.ac.jp [Laboratory of Plasma Physics and Engineering, Hokkaido University, Kita-13, Nishi-8, Kita-ku, Sapporo 060-8628 (Japan); Hino, Tomoaki; Yamauchi, Yuji; Nobuta, Yuji [Laboratory of Plasma Physics and Engineering, Hokkaido University, Kita-13, Nishi-8, Kita-ku, Sapporo 060-8628 (Japan); Akiba, Masato; Enoeda, Mikio [Japan Atomic Energy Agency, 801-1, Mukoyama, Naka 311-0193 (Japan)

    2013-10-15

    Lithium titanate (Li{sub 2}TiO{sub 3}) pebbles were irradiated with deuterium ions with energy of 1.7 keV and then exposed to helium or helium–hydrogen mixed gas at various temperatures, in order to evaluate the effects of gas exposure on deuterium removal from the pebbles. The amounts of residual deuterium in the pebbles were measured by thermal desorption spectroscopy. The mixing of hydrogen gas into helium gas enhanced the removal amount of deuterium. In other words, the amount of residual deuterium after the helium–hydrogen mixed gas exposure at lower temperature was lower than that after the helium gas exposure. In addition, we also evaluated the pebbles exposed to the helium gas with different hydrogen mixture ratio from 0% to 1%, at 573 K. Although the amount of residual deuterium in the pebbles after the exposure decreased with increasing the hydrogen mixture ratio, the implanted deuterium partly remained after the exposure. These results suggest that the tritium inventory may occur at low temperature region in the blanket during the operation.

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

    Silva, Alice Cunha da; Su, Jian

    2013-01-01

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

  12. Measurement of thermal expansion for a Li2TiO3 pebble bed

    Hisashi Tanigawa; Mikio Enoeda; Masato Akiba

    2006-01-01

    In the current design of the blanket with ceramic breeders, pebbles of breeding materials are packed into a container and used as a pebble bed. Thermal and mechanical conditions externally loaded on the bed affect thermal and mechanical properties of the bed. It is necessary to analyze thermo-mechanical properties of the bed under controlled thermal and mechanical conditions. In the present paper, thermal expansion of a Li 2 TiO 3 pebble bed was investigated. Our apparatus consists of a tensile test-apparatus and a measurement chamber. Pebbles of Li 2 TiO 3 with 2 mm diameter were used. They were packed into a container made of alumina. At first, thermal expansion of the apparatus was calibrated because the measured deformation included thermal expansions of the load rods and the container. Instead of the pebble bed, a column made of copper was installed and thermal expansion of the system was measured for the calibration. Taking into account the estimated thermal expansion of the column, thermal expansion of the rods and the container could be analyzed. Based on the correction, thermal expansion of the pebble bed was measured under compression of 0.1 MPa. Temperature of the bed was regulated from room temperature to 973 K. From the measured expansion of the bed, average thermal expansion coefficient was estimated. For the beds with different packing factors ranging from 65.5 to 68.5 %, thermal expansion coefficients were 1.4 ± 0. 10-5 K -1 . In the first measurement of the beds without pre-loading, expansion coefficients were larger for the cooling process than heating. When the beds were successively heated and cooled, the difference decreased. This means that relocation of the pebbles arises in the first heat treatment and progress of compaction is larger in the cooling process than heating. After a few heat treatments, packing states of the beds reach stable and expansion coefficients for both heat and cooling processes are close. In the case of the beds that

  13. Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 2: BOT helium cooled solid breeder blanket. Vol. 1

    Dalle Donne, M.; Boccaccini, L.V.; Bojarsky, E.; Deckers, H.; Dienst, W.; Doerr, L.; Fischer, U.; Giese, H.; Guenther, E.; Haefner, H.E.; Hofmann, P.; Kappler, F.; Knitter, R.; Kuechle, M.; Moellendorf, U. von; Norajitra, P.; Penzhorn, R.D.; Reimann, G.; Reiser, H.; Schulz, B.; Schumacher, G.; Schwenk-Ferrero, A.; Sordon, G.; Tsukiyama, T.; Wedemeyer, H.; Weimar, P.; Werle, H.; Wiegner, E.; Zimmermann, H.

    1991-10-01

    The BOT (Breeder Outside Tube) Helium Cooled Breeder Blanket for a fusion Demo reactor and the status of the R and D program is presented. This is the KfK contribution to the European Program for the Demo relevant test plankets to be irradiated in NET/ITER. Volume 1 (KfK 4928) contains the summary, volume 2 (KfK 4929) a more detailed version of the report. In both volumes are described the reasons for the selected design, the reference blanket design for the Demo reactor, the design of test blanket including the ancillary systems together with the present status of the relative R and D program in the fields of neutronic and thermohydraulic calculations, of the electromagnetic forces caused by disruptions, of the development and irradiation of the ceramic breeder material, of the tritium release and recovery, and of the technological investigations. An outlook is given on the required R and D program for the BOT Helium Cooled Solid Breeder Blanket prior to tests in NET/ITER and the proposed test program in NET/ITER. (orig.) [de

  14. Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 2: BOT helium cooled solid breeder blanket. Vol. 2

    Dalle Donne, M.; Boccaccini, L.V.; Bojarsky, E.; Deckers, H.; Dienst, W.; Doerr, L.; Fischer, U.; Giese, H.; Guenther, E.; Haefner, H.E.; Hofmann, P.; Kappler, F.; Knitter, R.; Kuechle, M.; Moellendorf, U. von; Norajitra, P.; Penzhorn, R.D.; Reimann, G.; Reiser, H.; Schulz, B.; Schumacher, G.; Schwenk-Ferrero, A.; Sordon, G.; Tsukiyama, T.; Wedemeyer, H.; Weimar, P.; Werle, H.; Wiegner, E.; Zimmermann, H.

    1991-10-01

    The BOT (Breeder Outside Tube) Helium Cooled Solid Breeder Blanket for a fusion Demo reactor and the status of the R and D program is presented. This is the KfK contribution to the European Program for the Demo relevant test blankets to be irradiated in NET/ITER. Volume 1 (KfK 4928) contains the summary, volume 2 (KfK 4929) a more detailed version of the report. In both volumes are described the reasons for the selected design, the reference blanket design for the Demo reactor, the design of the test blanket including the ancillary systems together with the present status of the relative R and D program in the fields of neutronic and thermohydraulic calculations, of the electromagnetic forces caused by disruptions, of the development and irradiation of the ceramic breeder material, of the tritium release and recovery, and of the technological investigations. An outlook is given on the required R and D program for the BOT Helium Cooled Solid Breeder Blanket prior to tests in NET/ITER and the proposed test program in NET/ITER. (orig.) [de

  15. Effective thermal conductivity of advanced ceramic breeder pebble beds

    Pupeschi, S., E-mail: simone.pupeschi@kit.edu; Knitter, R.; Kamlah, M.

    2017-03-15

    As the knowledge of the effective thermal conductivity of ceramic breeder pebble beds under fusion relevant conditions is essential for the development of solid breeder blanket concepts, the EU advanced and reference lithium orthosilicate material were investigated with a newly developed experimental setup based on the transient hot wire method. The effective thermal conductivity was investigated in the temperature range RT–700 °C. Experiments were performed in helium and air atmospheres in the pressure range 0.12–0.4 MPa (abs.) under a compressive load up to 6 MPa. Results show a negligible influence of the chemical composition of the solid material on the bed’s effective thermal conductivity. A severe reduction of the effective thermal conductivity was observed in air. In both atmospheres an increase of the effective thermal conductivity with the temperature was detected, while the influence of the compressive load was found to be small. A clear dependence of the effective thermal conductivity on the pressure of the filling gas was observed in helium in contrast to air, where the pressure dependence was drastically reduced.

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

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

    2014-01-01

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

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

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

    2009-01-01

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

  18. Photoassociation of cold metastable helium atoms

    Woestenenk, G.R.

    2001-01-01

    During the last decades the study of cold atoms has grown in a great measure. Research in this field has been made possible due to the development of laser cooling and trapping techniques. We use laser cooling to cool helium atoms down to a temperature of 1 mK and we are able to

  19. CFD investigation of thermal-hydraulic characteristics in a PBR core using different contact treatments between pebbles

    Ferng, Y.M.; Lin, K.Y.

    2014-01-01

    Highlights: • It is important to study thermal-hydraulic characteristics in a PBR for a HTGR. • A CFD model is proposed to simulate flow and heat transfer in a segment of pebbles. • Area and point contact treatments for adjacent pebbles are adopted in this study. • Predicted dependences of Nu and friction factor agree with the correlations. - Abstract: A high temperature gas cooled reactor (HTGR) with a pebble bed core (PBR) can be considered as one of the possible energy generation sources in the incoming future due to its inherently safe performance, lower power density, and higher conversion efficiency, etc. It is important to study the thermal-hydraulic characteristics in a PBR for optimum design and safe operation of a HTGR. In this paper, a computational fluid dynamics (CFD) methodology is proposed to investigate the thermal-hydraulic behavior in a segment of pebbles representing the central region of a PBR. Two kinds of contact modeling between adjacent pebbles are adopted, namely area and point contact treatments. The former contact treatment is a geometric approximation modeling. Based on the comparisons of thermal-hydraulic characteristics in the pebbles predicted by both contact treatments, no significant difference is revealed except for the near-wall secondary flow pattern. In addition, compared with the calculated results from the well-known correlations, the present predicted dependence of Nu number and friction factor on the particle Reynolds number shows good agreement qualitatively and quantitatively

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

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

    2015-01-01

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

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

    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

  2. Metaphysics methods development for high temperature gas cooled reactor analysis

    Seker, V.; Downar, T. J.

    2007-01-01

    Gas cooled reactors have been characterized as one of the most promising nuclear reactor concepts in the Generation-IV technology road map. Considerable research has been performed on the design and safety analysis of these reactors. However, the calculational tools being used to perform these analyses are not state-of-the-art and are not capable of performing detailed three-dimensional analyses. This paper presents the results of an effort to develop an improved thermal-hydraulic solver for the pebble bed type high temperature gas cooled reactors. The solution method is based on the porous medium approach and the momentum equation including the modified Ergun's resistance model for pebble bed is solved in three-dimensional geometry. The heat transfer in the pebble bed is modeled considering the local thermal non-equilibrium between the solid and gas, which results in two separate energy equations for each medium. The effective thermal conductivity of the pebble-bed can be calculated both from Zehner-Schluender and Robold correlations. Both the fluid flow and the heat transfer are modeled in three dimensional cylindrical coordinates and can be solved in steady-state and time dependent. The spatial discretization is performed using the finite volume method and the theta-method is used in the temporal discretization. A preliminary verification was performed by comparing the results with the experiments conducted at the SANA test facility. This facility is located at the Institute for Safety Research and Reactor Technology (ISR), Julich, Germany. Various experimental cases are modeled and good agreement in the gas and solid temperatures is observed. An on-going effort is to model the control rod ejection scenarios as described in the OECD/NEA/NSC PBMR-400 benchmark problem. In order to perform these analyses PARCS reactor simulator code will be coupled with the new thermal-hydraulic solver. Furthermore, some of the other anticipated accident scenarios in the benchmark

  3. Experimental study of flow field characteristics on bed configurations in the pebble bed reactor

    Jia, Xinlong; Gui, Nan; Yang, Xingtuan; Tu, Jiyuan; Jia, Haijun; Jiang, Shengyao

    2017-01-01

    Highlights: • PTV study of flow fields of pebble bed reactor with different configurations are carried out. • Some criteria are proposed to quantify vertical velocity field and flow uniformity. • The effect of different pebble bed configurations is also compared by the proposed criteria. • The displacement thickness is used analogically to analyze flow field characteristics. • The effect of mass flow variation in the stagnated region of the funnel flow is measured. - Abstract: The flow field characteristics are of fundamental importance in the design work of the pebble bed high temperature gas cooled reactor (HTGR). The different effects of bed configurations on the flow characteristics of pebble bed are studied through the PTV (Particle Tracking Velocimetry) experiment. Some criteria, e.g. flow uniformity (σ) and mass flow level (α), are proposed to estimate vertical velocity field and compare the bed configurations. The distribution of the Δθ (angle difference between the individual particle velocity and the velocity vector sum of all particles) is also used to estimate the resultant motion consistency level. Moreover, for each bed configuration, the thickness of displacement is analyzed to measure the effect of the funnel flow zone based on the boundary layer theory. Detailed information shows the quantified characteristics of bed configuration effects on flow uniformity and other characteristics; and the sequence of levels of each estimation criterion is obtained for all bed configurations. In addition, a good design of the pebble bed configuration is suggested and these estimation criteria can be also applied and adopted in testing other geometry designs of pebble bed.

  4. Performance characterization of the FLEX low pressure helium facility for fusion technology experiments

    Schlindwein, Georg, E-mail: schlindwein@kit.edu; Arbeiter, Frederik

    2014-10-15

    Highlights: • A gas loop for fusion R and D has been built and tested. • Facility requirements and their implementation are given. • The loop's functions and instrumentation are explained. • The loops performance has been characterized. - Abstract: FLEX (Fluid Dynamics Experimental Facility) is a multi-purpose small scale gas loop for research on fluid and thermodynamic investigations, especially heat transfer, flow field measurements and gas purification. Initially it was built for investigation on mini-channel gas-flow to design the HFTM module of IFMIF. Because of its versatility it offers a wide range of further applications, e.g. the research of pressure drops in mockups of breeder units of the helium cooled pebble bed (HCPB) test blanket module for ITER. The main parameters of the loop, which can be operated with inert gases and air are: (i) operation gas pressure 0.02–0.38 MPa abs., (ii) test section pressure head up to 0.12 MPa, (iii) tolerable gas temperature RT – 200 °C and (iv) mass flow rate 0.2–12 × 10{sup −3} kg/s for Helium. This paper gives a detailed view of the loop assembly with the components that generate and regulate the mass flow and loop pressure. The measurement instrumentation will be presented as well as a representative mass flow-pressure drop characteristic. Furthermore, the achievable gas purity will be discussed.

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

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

    2013-01-01

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

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

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

    2015-10-15

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

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

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

    2010-08-15

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

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

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

    2018-02-01

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

  9. Helium-Hydrogen Recovery System, Phase I

    National Aeronautics and Space Administration — Immense quantities of expensive liquefied helium are required at Stennis and Kennedy Space Centers for pre-cooling rocket engine propellant systems prior to filling...

  10. Study on restriction method for end-wall boundary layer thickness in axial helium gas compressor for gas turbine high temperature gas cooled reactor

    Takada, Shoji; Takizuka, Takakazu; Yan, Xing; Kunitomi, Kazuhiko; Inagaki, Yoshiyuki

    2009-01-01

    Aerodynamic performance test was carried out using a 1/3 scale, 4-stage model of the helium gas compressor to investigate an effect of end-wall over-camber to prevent decrease of axial velocity in the end-wall boundary layer. The model compressor consists of a rotor, 500 mm in diameter, which is driven by an electric motor at a rotational speed of 10800 rpm. The rotor blade span of the first stage is 34 mm. The test was carried out under the condition that the helium gas pressure of 0.88 MPa, temperature of 30degC, and mass flow rate of 12.47 kg/s at the inlet. A 3-dimensional aerodynamic code, which was verified using the test data, showed that axial velocity was lowered by using a blade which increased the inlet blade angle around the end-wall region of the casing side in comparison with that using the original design blade, because the inlet flow angle mismatched with the inlet blade angle of the rotor blade, as opposed to the prediction by a conventional air compressor design method. The overall adiabatic efficiency of the full scale 20-stage helium gas compressor was predicted 89.7% from the Reynolds number dependency of the test data by using the original design blade. (author)

  11. A CFD Study on Inlet Plenum Flow Field of Pebble Bed Reactor

    Kim, Min Hwan; Lee, Won Jae; Chang, Jong Hwa

    2005-01-01

    High temperature gas cooled reactor, largely divided into two types of PBR (Pebble Bed Reactor) and PMR (Prismatic Modular Reactor), has becomes great interest of researchers in connection with the hydrogen production. KAERI has started a project to develop the gas cooled reactor for the hydrogen production and has been doing in-depth study for selecting the reactor type between PBR and PMR. As a part of the study, PBMR (Pebble Bed Modular Reactor) was selected as a reference PBR reactor for the CFD analysis and the flow field of its inlet plenum was simulated with computational fluid dynamics program CFX5. Due to asymmetrical arrangement of pipes to the inlet plenum, non-uniform flow distribution has been expected to occur, giving rise to non-uniform power distribution at the core. Flow fields of different arrangement of inlet pipes were also investigated, as one of measures to reduce the non-uniformity

  12. Preliminary Safeguards Assessment for the Pebble-Bed Fluoride High-Temperature Reactor (PB-FHR) Concept

    Disser, Jay; Arthur, Edward; Lambert, Janine

    2016-09-01

    This report examines a preliminary design for a pebble bed fluoride salt-cooled high temperature reactor (PB-FHR) concept, assessing it from an international safeguards perspective. Safeguards features are defined, in a preliminary fashion, and suggestions are made for addressing further nuclear materials accountancy needs.

  13. Pebble Accretion in Turbulent Protoplanetary Disks

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

    2017-09-01

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

  14. Flow characteristics analysis of purge gas in unitary pebble beds by CFD simulation coupled with DEM geometry model for fusion blanket

    Chen, Youhua [University of Science and Technology of China, Hefei, Anhui, 230027 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Chen, Lei [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Luo, Guangnan [University of Science and Technology of China, Hefei, Anhui, 230027 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)

    2017-01-15

    Highlights: • A unitary pebble bed was built to analyze the flow characteristics of purge gas based on DEM-CFD method. • Flow characteristics between particles were clearly displayed. • Porosity distribution, velocity field distribution, pressure field distribution, pressure drop and the wall effects on velocity distribution were studied. - Abstract: Helium is used as the purge gas to sweep tritium out when it flows through the lithium ceramic and beryllium pebble beds in solid breeder blanket for fusion reactor. The flow characteristics of the purge gas will dominate the tritium sweep capability and tritium recovery system design. In this paper, a computational model for the unitary pebble bed was conducted using DEM-CFD method to study the purge gas flow characteristics in the bed, which include porosity distribution between pebbles, velocity field distribution, pressure field distribution, pressure drop as well as the wall effects on velocity distribution. Pebble bed porosity and velocity distribution with great fluctuations were found in the near-wall region and detailed flow characteristics between pebbles were displayed clearly. The results show that the numerical simulation model has an error with about 11% for estimating pressure drop when compared with the Ergun equation.

  15. Loads on pebble bed fuel elements

    Teuchert, E.; Maly, V.

    1974-03-15

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

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

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

    1979-10-01

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

  17. Tightly Coupled Multiphysics Algorithm for Pebble Bed Reactors

    Park, HyeongKae; Knoll, Dana; Gaston, Derek; Martineau, Richard

    2010-01-01

    We have developed a tightly coupled multiphysics simulation tool for the pebble-bed reactor (PBR) concept, a type of Very High-Temperature gas-cooled Reactor (VHTR). The simulation tool, PRONGHORN, takes advantages of the Multiphysics Object-Oriented Simulation Environment library, and is capable of solving multidimensional thermal-fluid and neutronics problems implicitly with a Newton-based approach. Expensive Jacobian matrix formation is alleviated via the Jacobian-free Newton-Krylov method, and physics-based preconditioning is applied to minimize Krylov iterations. Motivation for the work is provided via analysis and numerical experiments on simpler multiphysics reactor models. We then provide detail of the physical models and numerical methods in PRONGHORN. Finally, PRONGHORN's algorithmic capability is demonstrated on a number of PBR test cases.

  18. Development of a Novel Method for the Exploration of the Thermal Response of Superfluid Helium Cooled Superconducting Cables to Pulse Heat Loads

    Winkler, T.; Koettig, T.; van Weelderen, R.; Bremer, J.; ter Brake, H. J. M.

    Management of transient heat deposition in superconducting magnets and its extraction from the aforementioned is becoming increasingly important to bring high energy particle accelerator performance to higher beam energies and intensities. Precise knowledge of transient heat deposition phenomena in the magnet cables will permit to push the operation of these magnets as close as possible to their current sharing limit, without unduly provoking magnet quenches. With the prospect of operating the Large Hadron Collider at CERN at higher beam energies and intensities an investigation into the response to transient heat loads of LHC magnets, operating in pressurized superfluid helium, is being performed. The more frequently used approach mimics the cable geometry by resistive wires and uses Joule-heating to deposit energy. Instead, to approximate as closely as possible the real magnet conditions, a novel method for depositing heat in cable stacks made out of superconducting magnet-cables has been developed. The goal is to measure the temperature difference as a function of time between the cable stack and the superfluid helium bath depending on heat load and heat pulse length. The heat generation in the superconducting cable and precise measurement of small temperature differences are major challenges. The functional principle and experimental set-up are presented together with proof of principle measurements.

  19. Effect of carburizing helium environment on creep behavior of Ni-base heat-resistant alloys for high-temperature gas-cooled reactors

    Kurata, Yuji; Ogawa, Yutaka; Nakajima, Hajime

    1988-01-01

    Creep tests were conducted on Ni-base heat-resistant alloys Hastelloy XR and XR-II, i.e. versions of Hastelloy X modified for nuclear applications, at 950degC using four types of helium environment with different impurity compositions, and mainly the effect of carburization was examined. For all the materials tested, the values of creep rupture time obtained under the carburizing conditions were similar to or longer than those in the commonly used, standard test environment (JAERI Type B helium). The difference among the results was interpreted by the counterbalancing effects of the strengthening due to carburization and possible weakening caused under very low oxidizing potential. In the corrosion monitoring specimens pronounced carbon pick-up was observed in the environment with high carbon activity and very low oxidizing potential. Based on the results obtained in the present and the previous works, it is suggested that a moderate control of the impurity chemistry is important rather than simple purification of the coolant in protecting the material from the environment-enhanced degradation. Either condition with high or low extremes in the oxidizing and carburizing potentials may cause enhanced degradation and thus are desirable to be avoided at the elevated temperatures. (author)

  20. Fuel Development For Gas-Cooled Fast Reactors

    M. K. Meyer

    2006-06-01

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

  1. Geochemical fingerprints and pebbles zircon geochronology

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

  2. "Smart pebble" designs for sediment transport monitoring

    Valyrakis, Manousos; Alexakis, Athanasios; Pavlovskis, Edgars

    2015-04-01

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

  3. Test of a cryogenic helium pump

    Lue, J.W.; Miller, J.R.; Walstrom, P.L.; Herz, W.

    1981-01-01

    The design of a cryogenic helium pump for circulating liquid helium in a magnet and the design of a test loop for measuring the pump performance in terms of mass flow vs pump head at various pump speeds are described. A commercial cryogenic helium pump was tested successfully. Despite flaws in the demountable connections, the piston pump itself has performed satisfactorily. A helium pump of this type is suitable for the use of flowing supercritical helium through Internally Cooled Superconductor (ICS) magnets. It has pumped supercritical helium up to 7.5 atm with a pump head up to 2.8 atm. The maximum mass flow rate obtained was about 16 g/s. Performance of the pump was degraded at lower pumping speeds

  4. Gas-cooled fast reactor (GCFR) program review committee (PRC). Report No. 13 to members of Helium Breeder Associates (HBA), May 1979 -February 1980

    1980-01-01

    The responsiveness of the GCFR program structure to changes in design, schedule, and safety issues caused by the decisions to adopt the three-loop over the two-loop primary cooling loop design, to use upflow cooling, and to enhance the natural convection cooling of the plant in a shutdown mode was excellent. The PRC now feels that the criteria for design of a demonstration plant will be manageable and safe. Because of effort to resolve past issues, the PRC has not requested a cost breakdown at this time. It is felt that the HBA effort is properly directed in lieu of the current political malaise regarding future energy supplies and the breeder program in particular. The GCFR is responsive to change and could provide a commercial GCFR before the year 2000

  5. "Smart pebble" design for environmental monitoring applications

    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.

  6. Design of multi-input multi-output controller for magnetic bearing which suspends helium gas-turbine generator rotor for high temperature gas cooled reactor

    Takada, Shoji; Funatake, Yoshio; Inagaki, Yoshiyuki

    2009-01-01

    A design of a MIMO controller, which links magnetic forces of multiple magnetic bearings by feedback of multiple measurement values of vibration of a rotor, was proposed for the radial magnetic bearings for the generator rotor of helium gas turbine with a power output of 300 MWe. The generator rotor is a flexible rotor, which passes over the forth critical speed. A controller transfer function was derived at the forth critical speed, in which the bending vibration mode is similar to the one which is excited by unbalance mass to reduce a modeling error. A 1404-dimensional un-symmetric coefficient matrix of equation of state for the rotating rotor affected by Jayro effect was reduced by a modal decomposition using Schur decomposition to reduce a reduction error. The numerical results showed that unbalance response of rotor was 53 and 80 μm p-p , respectively, well below the allowable limits both at the rated and critical speeds. (author)

  7. Phenomenological studies and modelling of the gaseous impurities oxidation and adsorption mechanisms in helium: application for the purification system optimization in gas cooled nuclear reactors

    Legros, F.

    2008-01-01

    In GEN IV studies on future fission nuclear reactors, two concepts using helium as a coolant have been selected: GFR and VHTR. Among radioactive impurities and dusts, helium can contain H 2 , CO, CH 4 , CO 2 , H 2 O, O 2 , as well as nitrogenous species. To optimize the reactor functioning and lifespan, it is necessary to control the coolant chemical composition using a dedicated purification system. A pilot designed at the CEA allows studying this purification system. Its design includes three unit operations: H 2 and CO oxidation on CuO, then two adsorption steps. This study aims at providing a detailed analysis of the first and second purification steps, which have both been widely studied experimentally at laboratory scale. A first modelling based on a macroscopic approach was developed to represent the behaviour of the reactor and has shown that the CuO fixed bed conversion is dependent on the chemistry (mass transfer is not an issue) and is complete. The results of the structural analysis of the solids allow considering the CuO as particles made of 200 nm diameter grains. Hence, a new model at grain scale is proposed. It is highlighted that the kinetic constants from these two models are related with a scale factor which depends on geometry. A competition between carbon monoxide and hydrogen oxidation has been shown. Activation energies are around 30 kJ.mol-1. Simulation of the simultaneous oxidations leads to consider CO preferential adsorption. A similar methodology has been applied for CO 2 and H 2 O adsorption. The experimental isotherms showed a Langmuir type adsorption. Using this model, experimental and theoretical results agree. (author) [fr

  8. Experimental investigation on tritium release from lithium titanate pebble under high temperature of 1073 K

    Ochiai, Kentaro, E-mail: howartre@onid.oregonstate.edu [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki (Japan); Edao, Yuki; Kawamura, Yoshinori [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki (Japan); Hoshino, Tsuyoshi [Japan Atomic Energy Agency, Rokkasho-mura, Kamikita-gun, Aomori (Japan); Ohta, Masayuki; Sato, Satoshi; Konno, Chikara [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki (Japan)

    2015-10-15

    Highlights: • We have performed the tritium recovery experiment with the DT neutron source at 1073 K. • The tritium recovery corresponded with the calculated tritium production. • The chemical form of recovered tritium is affected by the temperature and kind of sweep gas. • The recovered HT increases at higher temperature and dry hydrogen circumstance. - Abstract: The temperature of Li{sub 2}TiO{sub 3} pebble breeder in a fusion DEMO blanket is assumed to be more than 1000 K. For the investigation of tritium release from a Li{sub 2}TiO{sub 3} pebble breeder blanket at such a high temperature, we have carried out a tritium release experiment with the DT neutron source at the JAEA-FNS. The Li{sub 2}TiO{sub 3} pebble (1.0–1.2 mm in diameter) of 70 g was put into a stainless steel container and installed into an assembly stratified with beryllium and Li{sub 2}TiO{sub 3} layers. During the DT neutron irradiation, the temperature was kept at 1073 K with wire heaters in the blanket container. Helium gas including 1% hydrogen gas (H{sub 2}/He) mainly flowed inside the container as the purge gas. Two chemical forms, HT and HTO, of extracted tritium were separately collected during the DT neutron irradiation by using water bubblers and CuO bed. The tritium activity in the water bubbler was measured by a liquid scintillation counter. To investigate the effect of moisture in the purge gas, we also performed the same experiments with H{sub 2}O/He gas (H{sub 2}O content: 1%) or pure helium gas. From our experiment at 1073 K, in the case of the purge gas includes H{sub 2}, it is indicated that the increasing tendency of HT release is similar to that of the dry H{sub 2}/He.

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

    1978-09-01

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

  10. Stability analysis of the high temperature thermal pebble bed nuclear reactor concept

    Vondy, D.R.

    1981-02-01

    A study was made of the stability of the high temperature gas-cooled pebble bed core against xenon-driven oscillation. This generic study indicated that a core as large as 3000 MW(t) could be stable. Several aspects present a challenge to analysis including the void space above the pebble bed, the effects of possible control rod configurations, and the temperature feedback contribution. Special methods of analysis were developed in this effort. Of considerable utility was the scheme of including an azimuthal buckling loss term in the neturon balance equations admitting direct solution of the first azimuthal harmonic for a core having azimuthal symmetry. This technique allows the linear stability analysis to be done solving two-dimensional (RZ) problems instead of three-dimensional problems. A scheme for removing the fundamental source contribution was also implemented to allow direct iteration toward the dominant harmonic solution, treating up to three dimensions with diffusion theory

  11. Measurement of the cooling capacity of an RMC-Cryosystems Model LTS 4.5-025 closed-cycle helium refrigerator

    De Zafra, R. L.; Mallison, W. H.; Emmons, L. K.; Koller, D.

    1991-01-01

    The cooling capacity of a recently purchased RMC-Cryosystems Model LTS 4.5-025 closed-cycle He refrigerator was measured over the range 4-35 K. It is found that the nominal cooling capacity of 250 mW is only met or exceeded over a narrow temperature range around 4.3 + or - 0.5 K, and that, above this range, there exists a considerable region of much lower cooling capacity, not exceeding about 100 mW. It is believed that this behavior results from use of a fixed-aperture Joule-Thompson expansion valve, and might be alleviated if the J-T valve could be adjusted to compensate for changing flow within the 5-20 K temperature range. Present performance may severely limit or prevent effective use in applications where an irreducible heat inflow exists which is greater than about 100 mW, yet substantially less than the quoted capacity at about 4 K.

  12. Real time thermal hydraulic model for high temperature gas-cooled reactor core

    Sui Zhe; Sun Jun; Ma Yuanle; Zhang Ruipeng

    2013-01-01

    A real-time thermal hydraulic model of the reactor core was described and integrated into the simulation system for the high temperature gas-cooled pebble bed reactor nuclear power plant, which was developed in the vPower platform, a new simulation environment for nuclear and fossil power plants. In the thermal hydraulic model, the helium flow paths were established by the flow network tools in order to obtain the flow rates and pressure distributions. Meanwhile, the heat structures, representing all the solid heat transfer elements in the pebble bed, graphite reflectors and carbon bricks, were connected by the heat transfer network in order to solve the temperature distributions in the reactor core. The flow network and heat transfer network were coupled and calculated in real time. Two steady states (100% and 50% full power) and two transients (inlet temperature step and flow step) were tested that the quantitative comparisons of the steady results with design data and qualitative analysis of the transients showed the good applicability of the present thermal hydraulic model. (authors)

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

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

    2006-01-01

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

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

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

    2014-10-15

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

  15. Risk-informed design of a pebble bed gas reactor

    Ritterbusch, Stanley; Dimitrijevic, Vesna; Simic Zdenko; Savkina Marina

    2003-01-01

    One of the major challenges to the successful deployment of new nuclear plants in the United States is the regulatory process, which is largely based on water-reactor design technology and operating experience. While ongoing and expected efforts to license new LWR designs are based primarily on current regulations, guidance, and past experience, the pre-application review of the gas-cooled Pebble Bed Modular Reactor (PBMR) has shown that efforts are being made to provide additional 'risk-informed' improvements to the licensing process. These improvements are aimed at resolving new design and regulatory issues using a plant-wide integrated evaluation method - state-of-the-art Probabilistic Risk Assessment - which addresses all significant design features and operating modes. The integrated PRA evaluation is supported by the usual deterministic design analyses, engineering judgments, and margins added to address uncertainties (i.e., defense-in-depth). The work performed for this paper was completed as part of the United States Department of Energy's Nuclear Energy Research Initiative. The purpose of this particular project was to develop the methods for a new 'highly risk-informed' design and regulatory process. In this work. PRA techniques were applied in order to provide an integrated and systematic analysis of the plant design, to quantify uncertainties and explicitly account for defense-in-depth features. This work concentrates on the application of the risk-informed principles to a new plant design such as the PBMR. The implementation example completed for this project included specification of the design configuration, use of the PRA to evaluate the design, and iterations to identify design changes that improve the overall level of safety and system reliability. This paper summarizes the new 'highly risk-informed' design process, the design of the PBMR, and the results obtained. These results, consistent with the known inherent safety features of a pebble

  16. Helium crystals

    Lipson, S.G.

    1987-01-01

    Hexagonal close-packed helium crystals in equilibrium with superfluid have been found to be one of the few systems in which an anisotropic solid comes into true thermodynamic equilibrium with its melt. The discovery of roughening transitions at the liquid-solid interface have shown this system to be ideal for the study of the statistical mechanics of interface structures. We describe the effect of roughening on the shape and growth of macroscopic crystals from both the theoretical and experimental points of view. (author)

  17. Quantum statistics and liquid helium 3 - helum 4 mixtures

    Cohen, E.G.D.

    1979-01-01

    The behaviour of liquid helium 3-helium 4 mixtures is considered from the point of view of manifestation of quantum statistics effects in macrophysics. The Boze=Einstein statistics is shown to be of great importance for understanding superfluid helium-4 properties whereas the Fermi-Dirac statistics is of importance for understanding helium-3 properties. Without taking into consideration the interaction between the helium atoms it is impossible to understand the basic properties of liquid helium 33 - helium 4 mixtures at constant pressure. Proposed is a simple model of the liquid helium 3-helium 4 mixture, namely the binary mixture consisting of solid spheres of two types subjecting to the Fermi-Dirac and Bose-Einstein statistics relatively. This model predicts correctly the most surprising peculiarities of phase diagrams of concentration dependence on temperature for helium solutions. In particular, the helium 4 Bose-Einstein statistics is responsible for the phase lamination of helium solutions at low temperatures. It starts in the peculiar critical point. The helium 4 Fermi-Dirac statistics results in incomplete phase lamination close to the absolute zero temperatures, that permits operation of a powerful cooling facility, namely refrigerating machine on helium solution

  18. Tritium release from beryllium pebbles after high temperature irradiation up to 3000 appm He in the HIDOBE-01 experiment

    Til, S. van, E-mail: vantil@nrg.eu [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Fedorov, A.V.; Stijkel, M.P.; Cobussen, H.L.; Mutnuru, R.K.; Idsert, P. van der [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)

    2013-11-15

    In the HIDOBE (HIgh DOse irradiation of BEryllium) irradiation program, various grades of constrained and unconstrained beryllium pebbles, beryllium pellets and titanium-beryllide samples are irradiated in the High Flux Reactor (HFR) in Petten at four different temperatures (between 698 K and 1023 K) for 649 days [1]. The first of two HIDOBE irradiation experiments, HIDOBE-01, was completed after achieving a DEMO relevant helium production level of 3000 appm and the samples are retrieved for postirradiation examination (PIE). This work shows preliminary results of the out-of-pile tritium release analysis performed on different grades of irradiated beryllium pebbles (different in size). Relationships between irradiation temperature, tritium inventory and microstructural evolution have been observed by light microscopy and scanning electron microscopy.

  19. Ceramic breeder pebble bed packing stability under cyclic loads

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

    2016-11-01

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

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

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

    2012-01-01

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

  1. Study on the conversion of H2 and CO from the helium carrier gas of high temperature gas-cooled reactor

    Liao Cuiping; Zheng Zhenhong; Shi Fuen

    1995-01-01

    The conversions of hydrogen and carbon monoxide into water vapor and carbon dioxide on CuO-ZnO-Al 2 O 3 catalyst are studied. The effects of different temperature, system atmospheric pressure, impurity gas concentration, flow and dew point on properties of cupric oxide bed are investigated. The conversion characteristics curves of H 2 and CO are given. Experimental data of conversion capacity, action period and conversion efficiency of CuO-ZnO-Al 2 O 3 are obtained and the optimal parameters are determined. The results show that the concentration of H 2 and CO of the effluent gas after purification can reach below 2 x 10 -6 , respectively. So it can meet the demands of high temperature gas-cooled reactor and also provide optimal design parameters and reliable data for conversion of H 2 and CO on CuO-ZnO-Al 2 O 3 catalyst

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

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

    2014-01-01

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

    Keppler, Istvan

    2013-01-01

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

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

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

    2013-05-15

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

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

    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.

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

    Huda, M.Q.; Obara, T.

    2008-01-01

    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

  7. Validation of CATHARE for gas-cooled reactors

    Fabrice Bentivoglio; Ola Widlund; Manuel Saez

    2005-01-01

    Full text of publication follows: Extensively validated and qualified for light-water reactor safety studies, the thermo-hydraulics code CATHARE has been adapted to deal also with gas-cooled reactor applications. In order to validate the code for these novel applications, CEA (Commissariat a l'Energie Atomique) has initiated an ambitious long-term experimental program. The foreseen experimental facilities range from small-scale loops for physical correlations, to component technology and system demonstration loops. In the short-term perspective, CATHARE is being validated against existing experimental data, in particular from the German power plant Oberhausen II and the South African Pebble-Bed Micro Model (PBMM). Oberhausen II, operated by the German utility EVO, is a 50 MW(e) direct-cycle Helium turbine plant. The power source is a gas burner rather than a nuclear reactor core, but the power conversion system resembles those of the GFR (Gas-cooled Fast Reactor) and other high-temperature reactor concepts. Oberhausen II was operated for more than 100 000 hours between 1974 and 1988. Design specifications, drawings and experimental data have been obtained through the European HTR project, offering a unique opportunity to validate CATHARE on a large-scale Brayton cycle. Available measurements of temperatures, pressures and mass flows throughout the circuit have allowed a very comprehensive thermohydraulic description of the plant, in steady-state conditions as well as during transients. The Pebble-Bed Micro Model (PBMM) is a small-scale model conceived to demonstrate the operability and control strategies of the South African PBMR concept. The model uses Nitrogen instead of Helium, and an electrical heater with a maximum rating of 420 kW. As the full-scale PBMR, the PBMM loop features three turbines and two compressors on the primary circuit, located on three separate shafts. The generator, however, is modelled by a third compressor on a separate circuit, with a

  8. Adsorption pump for helium pumping out

    Donde, A.L.; Semenenko, Yu.E.

    1981-01-01

    Adsorption pump with adsorbent cooling by liquid helium is described. Shuttered shield protecting adsorbent against radiation is cooled with evaporating helium passing along the coil positioned on the shield. The pump is also equipped with primed cylindrical shield, cooled with liquid nitrogen. The nitrogen shield has in the lower part the shuttered shield, on the pump casing there is a valve used for pump pre-burning, and valves for connection to recipient as well. Pumping- out rates are presented at different pressures and temperatures of adsorbent. The pumping-out rate according to air at absorbent cooling with liquid nitrogen constituted 5x10 -4 Pa-3000 l/s, at 2x10 -2 Pa-630 l/s. During the absorbent cooling with liquid hydrogen the pumping-out rate according to air was at 4x10 -4 Pa-580 l/s, at 2x10 -3 Pa-680 l/s, according to hydrogen - at 8x10 -5 Pa-2500 l/s, at 5x10 -3 Pa-4200 l/s. During adsorbent cooling with liquid helium the rate of pumping-out according to hydrogen at 3x10 5 Pa-2400% l/s, at 6x10 3 Pa-1200 l/s, and according to helium at 3.5x10 -5 Pa-2800 l/s, at 4x10 -3 Pa-1150 l/s. The limit vacuum is equal to 1x10 -7 Pa. The volume of the vessel with liquid helium is equal to 3.5 l. Helium consumption is 80 cm 3 /h. Consumption of liquid nitrogen from the shield is 400 cm 3 /h. The limit pressure in the pump is obtained after forevacuum pumping-out (adsorbent regeneration) at 300 K temperature. The pump is made of copper. The pump height together with primed tubes is 800 mm diameter-380 mm [ru

  9. A Pebble Bed Reactor cross section methodology

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

    2009-01-01

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

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

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

    2013-11-15

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

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

    Ho, Hai Quan; Obara, Toru

    2015-01-01

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

  12. Risk assessment of small-sized HTR with pebble-bed core

    Kroeger, W.; Mertens, J.; Wolters, J.

    1987-01-01

    Two recent concepts of small-sized HTR's (HTR-Modul and HTR-100) were analysed regarding their safety concepts and risk protection. In neither case do core cooling accidents contribute to the risk because of the low induced core temperatures. Water ingress accidents dominate the risk in both cases by detaching deposited fission products which can be released into the environment. For these accident sequences no early fatalities and practically no lethal case of cancer were computed. Both HTR concepts include adequate precautionary measures and an infinitely small risk according to the usual standards. The safety concepts make express use of the specific inherent safety features of pebble-bed HTR's. (orig.)

  13. Sustainability and Efficiency Improvements of Gas-Cooled High Temperature Reactors

    Marmier, Alain

    2012-01-01

    This thesis covers 3 fundamental aspects of High Temperature Reactor (HTR) performance: fuel testing under irradiation for maximized safety and sustainability, fuel architecture for improved economy and sustainability, and a novel Balance of Plant concept to enable future high-tech process heat applications with minimized R and D. The HTR concept features important inherent and passive safety characteristics: high thermal inertia and good thermal conductivity of the core; a negative Doppler coefficient; high quality of fuel elements and low power density. These features keep the core temperature within safe boundaries and minimise fission product release, even in case of severe accidents. The Very High Temperature reactor (VHTR) is based on the same safety concept as the initial HTR, but it aims at offering better economy with a higher reactor outlet temperature (and thus efficiency) and a high fuel discharge burn-up (and thus better sustainability). The inherent safety features of HTR have been demonstrated in small pebble-bed reactors in practice, but have to be replicated for reactors with industrially relevant size and power. An increase of the power density (in order to increase the helium coolant outlet temperature) leads to higher fuel temperatures and therefore higher fuel failure probability. The core of a pebble-bed reactor consists of 6 cm diameter spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. These pebbles contain thousands of 1 mm diameter fuel particles baked into a graphite matrix. These fuel particles, in turn, consist of a fuel kernel with successive coatings of pyrocarbon and silicon carbide layers. The coating layers are designed to contain the fission products that build up during operation of the reactor. The feasibility and performance of the fuel requires experimental verification in view of fuel qualification and licensing. For HTR fuel, the required test string comprises amongst others

  14. Impact of material system thermomechanics and thermofluid performance on He-cooled ceramic breeder blanket designs with SiCf/SiC

    Ying, Alice Y.; Yokomine, Takehiko; Shimizu, Akihiko; Abdou, Mohamed; Kohyama, Akira

    2004-01-01

    This paper presents results from a recent effort initiated under the JUPITER-II collaborative program for high temperature gas-cooled blanket systems using SiC f /SiC as a structural material. Current emphasis is to address issues associated with the function of the helium gas considered in the DREAM and ARIES-I concepts by performing thermomechanical and thermofluid analysis. The objective of the analysis is to guide future research focus for a task in the project. It is found that the DREAM concept has the advantage of achieving uniform temperature without threatening blanket pebble bed integrity by differential thermal stress. However, its superiority needs to be further justified by investigating the feasibility and economic issues involved in the tritium extraction technology

  15. Impact of material system thermomechanics and thermofluid performance on He-cooled ceramic breeder blanket designs with SiCf/SiC

    Ying, A.Y.; Abdou, M.; Yokomine, T.; Shimizu, A.; Kohyama, A.

    2008-01-01

    This paper presents results from a recent effort initiated under the JUPITER-II collaborative program for high temperature gas-cooled blanket systems using SiC/SiC as a structural material. Current emphasis is to address issues associated with the function of the helium gas considered in the DREAM and ARIES-I concepts by performing thermomechanical and thermofluid analysis. The objective of the analysis is to guide future research focus for a task in the project. It is found that the DREAM concept has the advantage of achieving uniform temperature without threatening blanket pebble bed integrity by differential thermal stress. However, its superiority needs to be further justified by investigating the feasibility and economic issues involved in the tritium extraction technology. (author)

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

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

    2004-01-01

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

  17. Transient analyses on the cooling channels of the DEMO HCPB blanket concept under accidental conditions

    Chen, Yuming; Ghidersa, Bradut-Eugen; Jin, Xue Zhou

    2016-01-01

    Highlights: • This paper presents transient CFD analyses on the cooling channels of the DEMO HCPB FW for accidental scenarios LOCA and LOFA. • In both LOCA & LOFA, the wall temperature increases quickly to an unacceptable level within seconds. • If the coolant flow rate is maintained at a half of nominal value in case of LOFA (partial LOFA), the wall temperature rises much slower, but will still leads to a damage of structure within minutes. • The simulated heat transfer coefficients were compared with empirical correlations. - Abstract: Helium Cooled Pebble Bed (HCPB) blanket concept is one of the DEMO (Demonstration Power Plant) blanket concepts running for the final DEMO design selection. In this paper, transient analyses on the cooling channels of the FW are carried out by means of CFD simulations for the selected accidental scenarios loss-of-coolant-accident (LOCA) and loss-of-flow-accident (LOFA). ANSYS-CFX is used for the simulations. The simulation results help to understand how fast the temperature of the FW can increase and what is the time window that is available until the temperature of the structural material reaches the design limit in order to be able to define a suitable protection strategy for the system. In view of later developments of the models, the heat transfer coefficients calculated with CFD are compared with the values predicted by two widely used correlations for turbulent pipe flows.

  18. Transient analyses on the cooling channels of the DEMO HCPB blanket concept under accidental conditions

    Chen, Yuming, E-mail: Yuming.chen@kit.edu; Ghidersa, Bradut-Eugen; Jin, Xue Zhou

    2016-11-01

    Highlights: • This paper presents transient CFD analyses on the cooling channels of the DEMO HCPB FW for accidental scenarios LOCA and LOFA. • In both LOCA & LOFA, the wall temperature increases quickly to an unacceptable level within seconds. • If the coolant flow rate is maintained at a half of nominal value in case of LOFA (partial LOFA), the wall temperature rises much slower, but will still leads to a damage of structure within minutes. • The simulated heat transfer coefficients were compared with empirical correlations. - Abstract: Helium Cooled Pebble Bed (HCPB) blanket concept is one of the DEMO (Demonstration Power Plant) blanket concepts running for the final DEMO design selection. In this paper, transient analyses on the cooling channels of the FW are carried out by means of CFD simulations for the selected accidental scenarios loss-of-coolant-accident (LOCA) and loss-of-flow-accident (LOFA). ANSYS-CFX is used for the simulations. The simulation results help to understand how fast the temperature of the FW can increase and what is the time window that is available until the temperature of the structural material reaches the design limit in order to be able to define a suitable protection strategy for the system. In view of later developments of the models, the heat transfer coefficients calculated with CFD are compared with the values predicted by two widely used correlations for turbulent pipe flows.

  19. Implementation of the superfluid helium phase transition using finite element modeling: Simulation of ransient heat transfer and He-I/He-II phase front movement in cooling channels of superconducting magnets

    Bielert, Erwin; Verweij, A.P.; ten Kate, Herman H.J.

    2013-01-01

    In the thermal design of high magnetic field superconducting accelerator magnets, the emphasis is on the use of superfluid helium as a coolant and stabilizing medium. The very high effective thermal conductivity of helium below the lambda transition temperature significantly helps to extract heat

  20. Mechanics of binary and polydisperse spherical pebble assembly

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

    2012-01-01

    The micromechanical behavior of an assembly of binary and polydisperse spherical pebbles is studied using discrete element method (DEM) accounting for microscopic interactions between individual pebbles. A in-house DEM code has been used to simulate the assemblies consisting of different pebble diameters and the results of the simulations are compared with that of mono-size pebble assemblies. The effect of relative radii and volume fraction of the pebbles on the macroscopic stress–strain response is discussed. Furthermore, the effect of packing factor and coefficient of friction on the overall stress–strain behavior of the system is studied in detail. The shear (tangential) stiffness between the particles is also another influencing parameter. For a very small shear stiffness the system shows a strong dependence on the packing factor while a pebble material dependent shear stiffness shows a rather moderate dependence on the packing factor. For a similar packing factor, the mono-size assembly shows a stiff behavior during loading compared to binary assembly. However, the simulations do not show a significant difference between the two behaviors in contrast to the observations made in the experiments. The discrepancy can be attributed to (i) probable difference in packing factors for mono-size and binary assemblies in the experiments, (ii) arbitrary friction coefficient in the current model and (iii) the tangential interaction (constant shear stiffness) implemented in the present model which needs further modification as a function of the load history on the pebbles. Evolution of other micromechanical characteristics such as coordination number, contact force distribution and stored elastic energy of individual pebbles as a function of external load and system parameters is presented which can be used to estimate important macroscopic properties such as overall thermal conductivity and crushing resistance of the pebble beds.

  1. Mechanics of binary and polydisperse spherical pebble assembly

    Annabattula, R.K., E-mail: ratna.annabattula@kit.edu [Institute for Applied Materials (IAM-WBM), Karlsruhe Institute of Technology (KIT), D-76344, Eggenstein-Leopoldshafen (Germany); Gan, Y., E-mail: yixiang.gan@sydney.edu.au [School of Civil Engineering, University of Sydney, 2006 NSW, Sydney (Australia); Kamlah, M., E-mail: marc.kamlah@kit.edu [Institute for Applied Materials (IAM-WBM), Karlsruhe Institute of Technology (KIT), D-76344, Eggenstein-Leopoldshafen (Germany)

    2012-08-15

    The micromechanical behavior of an assembly of binary and polydisperse spherical pebbles is studied using discrete element method (DEM) accounting for microscopic interactions between individual pebbles. A in-house DEM code has been used to simulate the assemblies consisting of different pebble diameters and the results of the simulations are compared with that of mono-size pebble assemblies. The effect of relative radii and volume fraction of the pebbles on the macroscopic stress-strain response is discussed. Furthermore, the effect of packing factor and coefficient of friction on the overall stress-strain behavior of the system is studied in detail. The shear (tangential) stiffness between the particles is also another influencing parameter. For a very small shear stiffness the system shows a strong dependence on the packing factor while a pebble material dependent shear stiffness shows a rather moderate dependence on the packing factor. For a similar packing factor, the mono-size assembly shows a stiff behavior during loading compared to binary assembly. However, the simulations do not show a significant difference between the two behaviors in contrast to the observations made in the experiments. The discrepancy can be attributed to (i) probable difference in packing factors for mono-size and binary assemblies in the experiments, (ii) arbitrary friction coefficient in the current model and (iii) the tangential interaction (constant shear stiffness) implemented in the present model which needs further modification as a function of the load history on the pebbles. Evolution of other micromechanical characteristics such as coordination number, contact force distribution and stored elastic energy of individual pebbles as a function of external load and system parameters is presented which can be used to estimate important macroscopic properties such as overall thermal conductivity and crushing resistance of the pebble beds.

  2. Charged condensate and helium dwarf stars

    Gabadadze, Gregory; Rosen, Rachel A, E-mail: gg32@nyu.edu, E-mail: rar339@nyu.edu [Center for Cosmology and Particle Physics, Department of Physics, New York University, New York, NY 10003 (United States)

    2008-10-15

    White dwarf stars composed of carbon, oxygen and heavier elements are expected to crystallize as they cool down below certain temperatures. Yet, simple arguments suggest that the helium white dwarf cores may not solidify, mostly because of zero-point oscillations of the helium ions that would dissolve the crystalline structure. We argue that the interior of the helium dwarfs may instead form a macroscopic quantum state in which the charged helium-4 nuclei are in a Bose-Einstein condensate, while the relativistic electrons form a neutralizing degenerate Fermi liquid. We discuss the electric charge screening, and the spectrum of this substance, showing that the bosonic long-wavelength fluctuations exhibit a mass gap. Hence, there is a suppression at low temperatures of the boson contribution to the specific heat-the latter being dominated by the specific heat of the electrons near the Fermi surface. This state of matter may have observational signatures.

  3. Analysis of a 115MW, 3 shaft, helium Brayton cycle

    Pradeepkumar, K.N.

    2002-01-01

    This research theme is originated from a development project that is going on in South Africa, for the design and construction of a closed cycle gas turbine plant using gas-cooled reactor as the heat source to generate 115 MW of electricity. South African Power utility company, Eskorn, promotes this developmental work through its subsidiary called PBMR (Pebble Bed Modular Reactor). Some of the attractive features of this plant are the inherent and passive safety features, modular geometry, small evacuation area, small infrastructure requirements for the installation and running of the plant, small construction time, quick starting and stopping and also low operational cost. This exercise is looking at the operational aspects of a closed cycle gas turbine, the finding of which will have a direct input towards the successful development and commissioning of the plant. A thorough understanding of the fluid dynamics in this three-shaft system and its transient performance analysis were the two main objectives of this research work. A computer programme called GTSI, developed by a previous Cranfield University research student, has been used in this as a base programme for the performance analysis. Some modifications were done on this programme to improve its control abilities. The areas covered in the performance analysis are Start-up, Shutdown and Load ramping. A detailed literature survey has been conducted to learn from the helium Turbo machinery experiences, though it is very limited. A critical analysis on the design philosophy of the PBMR is also carried out as part of this research work. The performance analysis has shown the advantage, disadvantage and impact of various power modulation methods suggested for the PBMR. It has tracked the effect of the operations of the various valves included in the PBMR design. The start-up using a hot gas injection has been analysed in detail and a successful start region has been mapped. A start-up procedure is also written

  4. Investigating effects of BCC and FCC arrangements on flow and heat transfer characteristics in pebbles through CFD methodology

    Ferng, Yuh Ming, E-mail: ymferng@ess.nthu.edu.tw [Department of Engineering and System Science, Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Sec. 2. Kuang-Fu Rd., Hsingchu 30013, Taiwan, ROC (China); Lin, Kun-Yueh [Department of Engineering and System Science, Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Sec. 2. Kuang-Fu Rd., Hsingchu 30013, Taiwan, ROC (China)

    2013-05-15

    Highlights: ► An HTGR would be one of the possible energy generation sources. ► We propose a CFD model to study effects of pebble arrangements for a PRB core. ► The entrance effect on the Nu number can be reasonably captured. ► The present predicted Nu versus Re{sub p} shows good agreement with data and correlation. ► Using FCC lattice in a core, simulation results may be non-conservative. -- Abstract: A high temperature gas cooled reactor (HTGR) would be one of the possible energy generation sources due to its advantages of inherently safety performance and higher conversion efficiency, etc. However, safety is the most important issue for its commercialization in energy industry. It is very crucial for safety design and operation of an HTGR to investigate its thermal–hydraulic characteristics. In this article, a computational fluid dynamics (CFD) methodology is proposed to investigate effects of different arrangements on these characteristics for an HTGR with a pebble bed (PB) core. Two kinds of arrangement: body-centered cubic (BCC) and face-centered cubic (FCC) are studies herein. Based on the simulation results, higher heat transfer capability and lower pebble temperature are predicted in the pebbles with the FCC-arrangement. The thermally fully-developed flow condition may be reached, which is shown in the result that the predicted average Nussel (Nu) number decreases from the 1st layer and reaches to an asymptotic value as the gas passes through the 6th layer of pebbles. This entrance effect reveals that the system codes using the correlations developed from the fully-developed flow condition can be appropriately applied in the entire PBR core. In addition, the present predicted dependence of Nu number on the inlet Reynolds (Re) number shows good agreement with that obtained from the well-known KTA. Measured data of Nu number versus Re number are also used to validate the CFD model.

  5. Conceptual design of helium experimental loop

    Yu Xingfu; Feng Kaiming

    2007-01-01

    In a future demonstration fusion power station (DEMO), helium is envisaged as coolant for plasma facing components, such as blanket and dive,or. All these components have a very complex geometry, with many parallel cooling channels, involving a complex helium flow distribution. Test blanket modules (TBM) of this concept will under go various tests in the experimental reactor ITER. For the qualification of TBM, it is indispensable to test mock-ups in a helium loop under realistic pressure and temperature profiles, in order to validate design codes, especially regarding mass flow and heat transition processes in narrow cooling channels. Similar testing must be performed for DEMO blanket, currently under development. A Helium Experimental Loop (HELOOP) is planed to be built for TBM tests. The design parameter of temperature, pressure, flow rate is 550 degree C, 10 MPa, l kg/s respectively. In particular, HELOOP is able to: perform full-scale tests of TBM under realistic conditions; test other components of the He-cooling system in ITER; qualify the purification circuit; obtain information for the design of the ITER cooling system. The main requirements and characteristics of the HELOOP facility and a preliminary conceptual design are described in the paper. (authors)

  6. Use of thorium for high temperature gas-cooled reactors

    Guimarães, Cláudio Q., E-mail: claudio_guimaraes@usp.br [Universidade de São Paulo (USP), SP (Brazil). Instituto de Física; Stefani, Giovanni L. de, E-mail: giovanni.stefani@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Santos, Thiago A. dos, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil)

    2017-07-01

    The HTGR ( High Temperature Gas-cooled Reactor) is a 4{sup th} generation nuclear reactor and is fuelled by a mixture of graphite and fuel-bearing microspheres. There are two competitive designs of this reactor type: The German “pebble bed” mode, which is a system that uses spherical fuel elements, containing a graphite-and-fuel mixture coated in a graphite shell; and the American version, whose fuel is loaded into precisely located graphite hexagonal prisms that interlock to create the core of the vessel. In both variants, the coolant consists of helium pressurised. The HTGR system operates most efficiently with the thorium fuel cycle, however, so relatively little development has been carried out in this country on that cycle for HTGRs. In the Nuclear Engineering Centre of IPEN (Instituto de Pesquisas Energéticas e Nucleares), a study group is being formed linked to thorium reactors, whose proposal is to investigate reactors using thorium for {sup 233}U production and rejects burning. The present work intends to show the use of thorium in HTGRs, their advantages and disadvantages and its feasibility. (author)

  7. Liquid helium target

    Fujii, Y.; Kitami, T.; Torikoshi, M.

    1984-12-01

    A liquid helium target system has been built and used for the experiment on the reaction 4 He(γ, p). The target system has worked satisfactorily; the consumption rate of liquid helium is 360 ml/h and the cryogenic system retains liquid helium for about ten hours. The structure, operation and performance of the target system are reported. (author)

  8. Matrix formulation of pebble circulation in the pebbed code

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

    2002-01-01

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

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

    Gomez A, R.; Espinosa P, G.

    2009-10-01

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

  10. Commercial helium reserves, continental rifting and volcanism

    Ballentine, C. J.; Barry, P. H.; Hillegonds, D.; Fontijn, K.; Bluett, J.; Abraham-James, T.; Danabalan, D.; Gluyas, J.; Brennwald, M. S.; Pluess, B.; Seneshens, D.; Sherwood Lollar, B.

    2017-12-01

    Helium has many industrial applications, but notably provides the unique cooling medium for superconducting magnets in medical MRI scanners and high energy beam lines. In 2013 the global supply chainfailed to meet demand causing significant concern - the `Liquid Helium Crisis' [1]. The 2017 closure of Quatar borders, a major helium supplier, is likely to further disrupt helium supply, and accentuates the urgent need to diversify supply. Helium is found in very few natural gas reservoirs that have focused 4He produced by the dispersed decay (a-particle) of U and Th in the crust. We show here, using the example of the Rukwa section of the Tanzanian East African Rift, how continental rifting and local volcanism provides the combination of processes required to generate helium reserves. The ancient continental crust provides the source of 4He. Rifting and associated magmatism provides the tectonic and thermal mechanism to mobilise deep fluid circulation, focusing flow to the near surface along major basement faults. Helium-rich springs in the Tanzanian Great Rift Valley were first identified in the 1950's[2]. The isotopic compositions and major element chemistry of the gases from springs and seeps are consistent with their release from the crystalline basement during rifting [3]. Within the Rukwa Rift Valley, helium seeps occur in the vicinity of trapping structures that have the potential to store significant reserves of helium [3]. Soil gas surveys over 6 prospective trapping structures (1m depth, n=1486) show helium anomalies in 5 out of the 6 at levels similar to those observed over a known helium-rich gas reservoir at 1200m depth (7% He - Harley Dome, Utah). Detailed macroseep gas compositions collected over two days (n=17) at one site allows us to distinguish shallow gas contributions and shows the deep gas to contain between 8-10% helium, significantly increasing resource estimates based on uncorrected values (1.8-4.2%)[2,3]. The remainder of the deep gas is

  11. Liquid-helium-cooled Michelson interferometer

    Augason, G. C.; Young, N.

    1972-01-01

    Interferometer serves as a rocket-flight spectrometer for examination of the far infrared emission spectra of astronomical objects. The double beam interferometer is readily adapted to make spectral scans and for use as a detector of discrete line emissions.

  12. Construction and testing of a double acting bellows liquid helium pump

    Burns, W.A.; Green, M.A.; Ross, R.R.; Van Slyke, H.

    1980-05-01

    The double acting reciprocating bellows liquid helium pump built and tested at the Lawrence Berkeley Laboratory is described. The pump is capable of delivering 50 gs -1 of liquid helium to supply the two-phase cooling sytem for a large superconducting magnet. The pump is driven by a torque motor at room temperature; the reciprocating motion is transmitted to the pump through a shaft which operates between room temperature and 4 0 K. The design details of this liquid helium pump are presented. The helium pump has operated in a helium bath and in pumped forced flow helium circuits. The results of these experimental tests are presented in this report

  13. ATLAS' major cooling project

    2005-01-01

    In 2005, a considerable effort has been put into commissioning the various units of ATLAS' complex cryogenic system. This is in preparation for the imminent cooling of some of the largest components of the detector in their final underground configuration. The liquid helium and nitrogen ATLAS refrigerators in USA 15. Cryogenics plays a vital role in operating massive detectors such as ATLAS. In many ways the liquefied argon, nitrogen and helium are the life-blood of the detector. ATLAS could not function without cryogens that will be constantly pumped via proximity systems to the superconducting magnets and subdetectors. In recent weeks compressors at the surface and underground refrigerators, dewars, pumps, linkages and all manner of other components related to the cryogenic system have been tested and commissioned. Fifty metres underground The helium and nitrogen refrigerators, installed inside the service cavern, are an important part of the ATLAS cryogenic system. Two independent helium refrigerators ...

  14. Neutron-induced helium implantation in GCFR cladding

    Yamada, H.; Poeppel, R.B.; Sevy, R.H.

    1980-10-01

    The neutron-induced implantation of helium atoms on the exterior surfaces of the cladding of a prototypic gas-cooled fast reactor (GCFR) has been investigated analytically. A flux of recoil helium particles as high as 4.2 x 10 10 He/cm 2 .s at the cladding surface has been calculated at the peak power location in the core of a 300-MWe GCFR. The calculated profile of the helium implantation rates indicates that although some helium is implanted as deep as 20 μm, more than 99% of helium particles are implanted in the first 2-μm-deep layer below the cladding surface. Therefore, the implanted helium particles should mainly affect surface properties of the GCFR cladding

  15. Lay-out of the He-cooled solid breeder model B in the European power plant conceptual study

    Hermsmeyer, S.; Malang, S.; Fischer, U.; Gordeev, S.

    2003-01-01

    The European helium cooled pebble bed (HCPB) blanket concept is the basis for one of two limited-extrapolation plant models that are being elaborated within the European power plant conceptual study (PPCS). In addition to addressing the case for fusion safety and environmental compatibility, following earlier studies like SEAFP or SEAL, this reactor study puts emphasis on plant availability and economic viability, which are closely related to specific plant models and require a detailed lay-out of the fusion power core and a consideration of the overall plant (balance of plant). Within the development of in-vessel components for the plant model, the major tasks to be carried out were: (i) adaptation of the HCPB concept--featuring separate pebble beds of ceramic breeder and Beryllium neutron multiplier and reduced-activation ferritic-martensitic steel EUROFER as structural material--to the large module segmentation chosen for reasons of plant availability in part II of the PPCS; (ii) proposal of a concept for a Helium cooled divertor compatible with a maximum of 10 MW/m 2 heat flux to satisfy the requirements of reasonably extrapolated plasma physics; (iii) lay-out of the major plant model components and integration into the in-vessel dimensions found from system code calculations for a power plant of 1500 MW electrical output and iterated data on the plant model performance. The paper defines all major in-vessel components of plant model B, as it is called in the PPCS, namely (i) the unit of FW, blanket and high temperature shield that is to be replaced regularly; (ii) the low temperature shield that is laid out as a lifetime component of the reactor; (iii) the divertor; and (iv) the in-vessel manifolding. Results are presented for the thermal-hydraulic performance of the components and for the thermal-mechanical behaviour of the blanket and the divertor target plate. These results suggest, together with results from the wider exploration of the plant model within

  16. Liquid helium plant in Dubna

    Agapov, N.N.; Baldin, A.M.; Kovalenko, A.D.

    1995-01-01

    The liquid-helium cooling capacity installed at the Laboratory of High Energies is about 5 kw at a 4.5 K temperature level. It is provided with four industrial helium liquefiers of 1.6 kw/4.5 K each. They have been made by the Russian enterprise NPO GELYMASH and upgraded by the specialists of the Laboratory. The first one was put into operation in 1980, the two others in 1991, and the last one is under commissioning. The development of the LHE cryoplant was concerned with the construction of the new superconducting accelerator Nuclotron aimed to accelerate nuclei and heavy ions up to energies of 6 GeV/u. The first test run at the Nuclotron was carried out in March 1993, and the total running time has been about 2000 hours up to now. Since 1992 the cryoplant has been intensively used by the users outside the Laboratory. More than a million liters of liquid helium was provided in 1993 for such users. The reliability of the cryoplant system was as high as 98 percent for 4500 hours of operation in 1993-1994. 7 refs., 4 figs., 1 tab

  17. The CEPHEID Project - Conceptual and Feasibility Study

    Benoit, Ph.

    1997-02-01

    Different blanket concepts have been proposed for future thermonuclear fusion reactors. Among the two candidate concepts selected, the Helium Cooled Pebble Bed (HCPB), which is based on ceramic pebble beds, lithium orthosilicate and beryllium, cooled by helium. The paper describes an experimental device to be installed in the BR-2 reactor and aiming at operating a HCPB test module in nominal conditions, to the maximum possible extent. The device has been called CEramic Pebble bed Helium cooled Irradiation for Demo (CEPHEID)

  18. The CEPHEID Project - Conceptual and Feasibility Study

    Benoit, Ph.

    1997-02-01

    Different blanket concepts have been proposed for future thermonuclear fusion reactors. Among the two candidate concepts selected, the Helium Cooled Pebble Bed (HCPB), which is based on ceramic pebble beds, lithium orthosilicate and beryllium, cooled by helium. The paper describes an experimental device to be installed in the BR-2 reactor and aiming at operating a HCPB test module in nominal conditions, to the maximum possible extent. The device has been called CEramic Pebble bed Helium cooled Irradiation for Demo (CEPHEID).

  19. Rotary magnetic refrigerator for superfluid helium production

    Hakuraku, Y.; Ogata, H.

    1986-01-01

    A new rotary-magnetic refrigerator designed to obtain superfluid helium temperatures by executing a magnetic Carnot cycle is developed. A rotor containing 12 magnetic refrigerants (gadolinium-gallium-garnet) is immersed in liquid helium at 4.2 K and rotated at constant speed in a steady magnetic field distribution. Performance tests demonstrate that the new rotary refrigerator is capable of obtaining a temperature of 1.48 K. The maximum useful cooling power obtained at 1.8 K is 1.81 W which corresponds to a refrigeration efficiency of 34%

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

    Hermsmeyer, S.; Reimann, J.

    2002-01-01

    Pebble beds of ceramic material are investigated within the framework of developing solid breeder blankets for future fusion power plants. A thermo-mechanical characterisation of such pebble beds is mandatory for understanding the behaviour of pebble beds, and thus the overall blanket, under fusion environment conditions. The mechanical behaviour of pebble beds is typically explored with uni-axial, bi-axial and tri-axial compression experiments. The latter two types of experiment are particularly revealing since they contain explicitly, beyond a compression behaviour of the bed, information on the conditions for pebble flow, i.e. macroscopic relocation, in the pebble bed. (orig.)

  1. Neutronic modeling of pebble bed reactors in APOLLO2

    Grimod, M.

    2010-01-01

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

  2. Dynamics and inherent safety features of small modular high temperature gas-cooled reactors

    Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

    1986-01-01

    Investigations were made at Oak Ridge National Laboratory to characterize the dynamics and inherent safety features of various modular high temperature gas-cooled reactor (HTGR) designs. This work was sponsored by the US Nuclear Regulatory Commission's HTGR Safety Research program. The US Department of Energy (DOE) and the Gas Cooled Reactor Associates (GCRA) have sponsored studies of several modular HTGR concepts, each having it own unique advantageous economic and inherent safety features. The DOE design team has recently choses a 350-MW(t) annular core with prismatic, graphite matrix fuel for its reference plant. The various safety features of this plant and of the pebble-bed core designs similar to those currently being developed and operated in the Federal Republic of Germany (FRG) are described. A varity of postulated accident sequences involving combinations of loss of forced circulation of the helium primary coolant, loss of primary coolant pressurization, and loss of normal and backup heat sinks were studied and are discussed. Results demonstrate that each concept can withstand an uncontrolled heatup accident without reaching excessive peak fuel temperatures. Comparisons of calculated and measured response for a loss of forced circulation test on the FRG reactor, AVR, are also presented. 10 refs

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

    Kim, Hong Chul

    2012-02-01

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

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

    Chen, Leisheng; Lee, Wooram; Lee, Jaeyoung

    2016-01-01

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

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

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

    2016-05-15

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

  6. Two phase cooling for superconducting magnets

    Eberhard, P.H.; Gibson, G.A.; Green, M.A.; Ross, R.R.; Smits, R.G.

    1986-01-01

    Comments on the use of two phase helium in a closed circuit tubular cooling system and some results obtained with the TPC superconducting magnet are given. Theoretical arguments and experimental evidence are given against a previously suggested method to determine helium two phase flow regimes. Two methods to reduce pressure in the magnet cooling tubes during quenches are discussed; 1) lowering the density of helium in the magnet cooling tubes and 2) proper location of pressure relief valves. Some techniques used to protect the refrigerator from too much cold return gas are also mentioned

  7. Two phase cooling for superconducting magnets

    Eberhard, P.H.; Gibson, G.A.; Green, M.A.; Ross, R.R.; Smits, R.G.; Taylor, J.D.; Watt, R.D.

    1986-01-01

    Comments on the use of two phase helium in a closed circuit tubular cooling system and some results obtained with the TPC superconducting magnet are given. Theoretical arguments and experimental evidence are given against a previously suggested method to determine helium two phase flow regimes. Two methods to reduce pressure in the magnet cooling tubes during quenches are discussed; (1) lowering the density of helium in the magnet cooling tubes and (2) proper location of pressure relief valves. Some techniques used to protect the refrigerator from too much cold return gas are also mentioned. 10 refs., 1 fig., 5 tabs

  8. Helium storage and control system for the PBMR

    Verkerk, E.C.

    1997-01-01

    The power conversion unit will convert the heat energy in the reactor core to electrical power. The direct-closed cycle recuperated Brayton Cycle employed for this concept consists of a primary helium cycle with helium powered turbo compressors and a power turbine. The helium is actively cooled with water before the compression stages. A recuperator is used to preheat the helium before entering the core. The start of the direct cycle is initiated by a mass flow from the helium inventory and control system via a jet pump. When the PBMR is connected to the grid, changes in power demand can be followed by changing the helium flow and pressure inside the primary loop. Small rapid adjustments can be performed without changing the helium inventory of the primary loop. The stator blade settings on the turbines and compressors are adjustable and it is possible to bypass reactor and turbine. This temporarily reduces the efficiency at which the power conversion unit is operating. Larger or long term adjustments require storage or addition of helium in order to maintain a sufficient level of efficiency in the power conversion unit. The helium will be temporarily stored in high pressure tanks. After a rise in power demand it will be injected back into the system. Some possibilities how to store the helium are presented in this paper. The change of helium inventory will cause transients in the primary helium loop in order to acquire the desired power level. At this stage, it seems that the change of helium inventory does not strongly effect the stability of the power conversion unit. (author)

  9. Investigation of impurity-helium solid phase decomposition

    Boltnev, R.E.; Gordon, E.B.; Krushinskaya, I.N.; Martynenko, M.V.; Pel'menev, A.A.; Popov, E.A.; Khmelenko, V.V.; Shestakov, A.F.

    1997-01-01

    The element composition of the impurity-helium solid phase (IHSP), grown by injecting helium gas jet, involving Ne, Ar, Kr, and Xe atoms and N 2 molecules, into superfluid helium, has been studied. The measured stoichiometric ratios, S = N H e / N I m, are well over the values expected from the model of frozen together monolayer helium clusters. The theoretical possibility for the freezing of two layers helium clusters is justified in the context of the model of IHSP helium subsystem, filled the space between rigid impurity centers. The process of decomposition of impurity-helium (IH)-samples taken out of liquid helium in the temperature range 1,5 - 12 K and the pressure range 10-500 Torr has been studied. It is found that there are two stages of samples decomposition: a slow stage characterized by sample self cooling and a fast one accompanied by heat release. These results suggest, that the IHSP consists of two types of helium - weakly bound and strongly bound helium - that can be assigned to the second and the first coordination helium spheres, respectively, formed around heavy impurity particles. A tendency for enhancement of IHSP thermo stability with increasing the impurity mass is observed. Increase of helium vapor pressure above the sample causes the improvement of IH sample stability. Upon destruction of IH samples, containing nitrogen atoms, a thermoluminescence induced by atom recombination has been detected in the temperature region 3-4,5 K. This suggests that numerous chemical reactions may be realized in solidified helium

  10. Trial fabrication of Be12Ti electrode for pebble production by rotating electrode method

    Uda, M.; Iwadachi, T.; Uchida, M.; Nakamichi, M.; Kawamura, H.

    2004-01-01

    Be 12 Ti has been one of candidates for advanced neutron multipliers, due to its high melting points and good chemical stability. Although Be 12 Ti is too brittle to product pebbles for neutron multipliers with the rotating electrode method (REM), a preliminary production of the pebbles which was made of two phase material of α-beryllium (α-Be) and Be 12 Ti was successfully demonstrated with REM. In this study a trial fabrication of the Be-5at%Ti'' ingot (α-Be + Be 12 Ti) for the REM electrode was carried out with a vacuum casting process. Three kinds of refractory crucibles (MgO, CaO and BeO) were tested for the evaluation of durability to the melt of Be-5%Ti. The water-cooled copper mould was applied for the casting mold to assist a one-direction solidification. The appearance (crack(s), shrinkage, etc.), microstructure and chemical analysis of the ''Be-5at%Ti'' ingots were investigated. As for the results of the trial fabrication, it was made clear that BeO crucible is most useful for the melting of Be-5at%Ti. The ingot, which was a size of φ85 x h 150 mm, had h 40mm sound portion from the bottom. The microstructure of the ingots showed two phases (probably α-Be and Be 12 Ti). The chemical composition of the Ti in the ingot were 3.1 - 6.2at%. (author)

  11. Li2O-pebble type tritium breeding blanket for fusion experimental reactor, 1

    Tone, Tatsuzo; Iida, Hiromasa; Tanaka, Yoshihisa

    1984-01-01

    The fusion experimental reactor is the next stage device in Japan, which is planned to be constructed following the critical plasma experimental device JT-60 being constructed at present. The breeding blanket installed in nuclear fusion reactors is one of most important structures, and it is required to satisfy the fundamental performance of producing and continuously recovering tritium as the nuclear fusion fuel, and other requirement in good coordination. The Li 2 O pebble type breeding blanket that Kawasaki Heavy Industries Ltd. has examined is the concept for resolving the problems of the mass transfer and thermal stress cracking of Li 2 O, which are important in blanket design. In this paper, the concept and characteristics of this breeding blanket are discussed from the viewpoint of the breeding and continuous recovery of tritium, the ease of manufacture and the maintenance of soundness. The breeding blanket is composed of breeding region, tritium purge region, cooling region, plasma stabilizing conductors and blanket container. Li 2 O is excellent in its tritium breeding performance and heat conductivity. The functions required for the breeding blanket, the fundamental structure, the examples of breeding blanket concept, the selection of breeding blanket concept, the characteristics of Li 2 O pebble type blanket and its future prospect are described. (Kako, I.)

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

    Wei Chunlin; Zhao Jing; Zhang Jian; Xia Bing

    2014-01-01

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

  13. COOLING STAGES OF CRYOGENIC SYSTEMS

    Троценко, А. В.

    2011-01-01

    The formalized definition for cooling stage of low temperature system is done. Based on existing information about the known cryogenic unit cycles the possible types of cooling stages are single out. From analyses of these stages their classification by various characteristics is suggested. The results of thermodynamic optimization of final throttle stage of cooling, which are used as working fluids helium, hydrogen and nitrogen, are shown.

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

    Brian Boer; Abderrafi M. Ougouag

    2011-03-01

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

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

    Boer, Brian; Ougouag, Abderrafi M.

    2011-01-01

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

  16. Current status and results of the PBMR -Pebble Box- benchmark within the framework of the IAEA CRP5 - 341

    Reitsma, F.; Tyobeka, B.

    2010-01-01

    The verification and validation of computer codes used in the analysis of high temperature gas cooled pebble bed reactor systems has not been an easy goal to achieve. A limited amount of tests and operating reactor measurements are available. Code-to-code comparisons for realistic pebble bed reactor designs often exhibit differences that are difficult to explain and are often blamed on the complexity of the core models or the variety of analysis methods and cross section data sets employed. For this reason, within the framework of the IAEA CRP5, the 'Pebble Box' benchmark was formulated as a simple way to compare various treatments of neutronics phenomena. The problem is comprised of six test cases which were designed to investigate the treatments and effects of leakage and heterogeneity. This paper presents the preliminary results of the benchmark exercise as received during the CRP and suggests possible future steps towards the resolution of discrepancies between the results. Although few participants took part in the benchmarking exercise, the results presented here show that there is still a need for further evaluation and in-depth understanding in order to build the confidence that all the different methods, codes and cross-section data sets have the capability to handle the various neutronics effects for such systems. (authors)

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

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

    2006-01-01

    High temperature gas-cooled reactors(HTGR) have been reviewed as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor(PBR) and a prismatic modular reactor(PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both a PBR and a PMR. The objective of this study is to predict the hot-spot fuel temperature distributions in a PBR and a PMR at a steady state. The computational fluid dynamics(CFD) code, CFX-10 is used to perform the three-dimensional analysis. The latest design data was used here based on the reference reactor designs, PBMR400 and GTMHR60

  18. Letters initiating Clean Water Act 404(c) review of mining at Pebble deposit

    Correspondence between EPA and the Pebble Limited Partnership and the State of Alaska initiating review under section 404(c) of the Clean Water Act of potential adverse environmental effects associated with mining the Pebble deposit in southwest Alaska.

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

    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.

  20. Tritium adsorption/release behaviour of advanced EU breeder pebbles

    Kolb, Matthias H.H., E-mail: matthias.kolb@kit.edu; Rolli, Rolf; Knitter, Regina

    2017-06-15

    The tritium loading of current grades of advanced ceramic breeder pebbles with three different lithium orthosilicate (LOS)/lithium metatitanate (LMT) compositions (20–30 mol% LMT in LOS) and pebbles of EU reference material, was performed in a consistent way. The temperature dependent release of the introduced tritium was subsequently investigated by temperature programmed desorption (TPD) experiments to gain insight into the desorption characteristics. The obtained TPD data was decomposed into individual release mechanisms according to well-established desorption kinetics. The analysis showed that the pebble composition of the tested samples does not severely change the release behaviour. Yet, an increased content of lithium metatitanate leads to additional desorption peaks at medium temperatures. The majority of tritium is released by high temperature release mechanisms of chemisorbed tritium, while the release of physisorbed tritium is marginal in comparison. The results allow valuable projections for the tritium release behaviour in a fusion blanket.

  1. Tritium adsorption/release behaviour of advanced EU breeder pebbles

    Kolb, Matthias H. H.; Rolli, Rolf; Knitter, Regina

    2017-06-01

    The tritium loading of current grades of advanced ceramic breeder pebbles with three different lithium orthosilicate (LOS)/lithium metatitanate (LMT) compositions (20-30 mol% LMT in LOS) and pebbles of EU reference material, was performed in a consistent way. The temperature dependent release of the introduced tritium was subsequently investigated by temperature programmed desorption (TPD) experiments to gain insight into the desorption characteristics. The obtained TPD data was decomposed into individual release mechanisms according to well-established desorption kinetics. The analysis showed that the pebble composition of the tested samples does not severely change the release behaviour. Yet, an increased content of lithium metatitanate leads to additional desorption peaks at medium temperatures. The majority of tritium is released by high temperature release mechanisms of chemisorbed tritium, while the release of physisorbed tritium is marginal in comparison. The results allow valuable projections for the tritium release behaviour in a fusion blanket.

  2. A virtual pebble game to ensemble average graph rigidity.

    González, Luis C; Wang, Hui; Livesay, Dennis R; Jacobs, Donald J

    2015-01-01

    The body-bar Pebble Game (PG) algorithm is commonly used to calculate network rigidity properties in proteins and polymeric materials. To account for fluctuating interactions such as hydrogen bonds, an ensemble of constraint topologies are sampled, and average network properties are obtained by averaging PG characterizations. At a simpler level of sophistication, Maxwell constraint counting (MCC) provides a rigorous lower bound for the number of internal degrees of freedom (DOF) within a body-bar network, and it is commonly employed to test if a molecular structure is globally under-constrained or over-constrained. MCC is a mean field approximation (MFA) that ignores spatial fluctuations of distance constraints by replacing the actual molecular structure by an effective medium that has distance constraints globally distributed with perfect uniform density. The Virtual Pebble Game (VPG) algorithm is a MFA that retains spatial inhomogeneity in the density of constraints on all length scales. Network fluctuations due to distance constraints that may be present or absent based on binary random dynamic variables are suppressed by replacing all possible constraint topology realizations with the probabilities that distance constraints are present. The VPG algorithm is isomorphic to the PG algorithm, where integers for counting "pebbles" placed on vertices or edges in the PG map to real numbers representing the probability to find a pebble. In the VPG, edges are assigned pebble capacities, and pebble movements become a continuous flow of probability within the network. Comparisons between the VPG and average PG results over a test set of proteins and disordered lattices demonstrate the VPG quantitatively estimates the ensemble average PG results well. The VPG performs about 20% faster than one PG, and it provides a pragmatic alternative to averaging PG rigidity characteristics over an ensemble of constraint topologies. The utility of the VPG falls in between the most

  3. Helium the disappearing element

    Sears, Wheeler M

    2015-01-01

    The subject of the book is helium, the element, and its use in myriad applications including MRI machines, particle accelerators, space telescopes, and of course balloons and blimps. It was at the birth of our Universe, or the Big Bang, where the majority of cosmic helium was created; and stellar helium production continues. Although helium is the second most abundant element in the Universe, it is actually quite rare here on Earth and only exists because of radioactive elements deep within the Earth. This book includes a detailed history of the discovery of helium, of the commercial industry built around it, how the helium we actually encounter is produced within the Earth, and the state of the helium industry today. The gas that most people associate with birthday party balloons is running out. “Who cares?” you might ask. Well, without helium, MRI machines could not function, rockets could not go into space, particle accelerators such as those used by CERN could not operate, fiber optic cables would not...

  4. Helium dilution refrigerator

    1973-01-01

    A new system of continuous heat exchange for a helium dilution refrigerator is proposed. The 3 He effluent tube is concurrent with the affluent mixed helium tube in a vertical downward direction. Heat exchange efficiency is enhanced by placing in series a number of elements with an enlarged surface area

  5. Helium localisation in tritides

    Flament, J.L.; Lozes, G.

    1982-06-01

    Study of titanium and LaNi 5 type alloys tritides lattice parameters evolution revealed that helium created by tritium decay remains in interstitial sites up to a limit material dependant concentration. Beyond this one exceeding helium precipites in voids [fr

  6. Consideration of heat transfer performance of helium-gas/water coolers in HENDEL

    Inagaki, Yoshiyuki; Miyamoto, Yoshiaki

    1986-10-01

    The helium engineering loop (HENDEL) has four helium-gas/water coolers, where the cooling water flows in the tubes and the helium gas flows on the shell side. Their cooling performance depends on mainly the heat transfer of helium gas on the shell side. This report describes the operational data of the coolers and the consideration of the heat transfer performance which is important for the design of coolers. It becomes clear that Donohue's equation is close to the operational data and conservative for the segmental baffle type cooler and preduction by Fishenden-Saunders or Zukauskas' equation is conservation for the step-up baffle type cooler. (author)

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

    1978-09-01

    This report describes and evaluates several gas-cooled reactor plant concepts under development within the Federal Republic of Germany (FRG). The concepts, based upon the use of a proven Pebble Bed Reactor (PBR) fuel element design, include nuclear heat generation for chemical processes and electrical power generation. Processes under consideration for the nuclear process heat plant (PNP) include hydrogasification of coal, steam gasification of coal, combined process, and long-distance chemical heat transportation. The electric plant emphasized in the report is the steam turbine cycle (HTR-K), although the gas turbine cycle (HHT) is also discussed. The study is a detailed description and evaluation of the nuclear portion of the various plants. The general conclusions are that the PBR technology is sound and that the HTR-K and PNP plant concepts appear to be achievable through appropriate continuing development programs, most of which are either under way or planned

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

    Saphier, D.

    1984-04-01

    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.)

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

    Moormann, R.

    1986-01-01

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

  10. Development of computational methods for the safety assessment of gas-cooled high-temperature and supercritical light-water reactors. Final report; Rechenmethoden zur Bewertung der Sicherheit von gasgekuehlten Hochtemperaturreaktoren und superkritischen Leichtwasserreaktoren. Abschlussbericht

    Buchholz, S.; Cron, D. von der; Hristov, H.; Lerchl, G.; Papukchiev, A.; Seubert, A.; Sureda, A.; Weis, J.; Weyermann, F.

    2012-12-15

    . Performed test and validation calculations for short and long term transients like withdrawal and ejection of all or single control rods, cold helium ingress or depressurized loss of forced cooling (DLOFC) demonstrate the applicability of TORT-TD/ATTICA3D to 3-D analyses of pebble bed HTR. Chapter 6 documents the extension made in ATHLET regarding application to supercritical water reactors. This includes the implementation of supercritical water as a working fluid and extensions of the model equations for the physics of heat transfer and pressure drop at supercritical water pressure as well as the extension of the material properties package to pressures above the critical point and the modeling of supercritical discharge. The extensions in ATHLET to simulate pebble-bed HTR are described in chapter 7. In ATHLET, the coolant helium has been implemented both as gas component and a working fluid. The material properties package has been properly extended. For the thermal hydraulic modeling of the reactor pressure vessel, a generic parallel channel model including cross connections has been developed for the PBMR-400 design. The HECU model in ATHLET has been extended to spherical geometries in order to simulate the heat transfer processes in HTR fuel pebbles with detailed representation of the TRISO particle fuel. In addition, ATHLET models of gas turbine and compressor have been developed and tested. Finally, chapter 8 documents the development and validation of ANSYS CFX for application to alternative reactor concepts. This includes extensions and applications of the CFX code regarding HPLWR requirements. Accuracy demonstrations of ANSYS CFX models for heat transfer and wall interfaces of gas cooled systems have been performed for several turbulence models by comparing with experimental data. Finally, the development and validation of the coupled code system ATHLET/ANSYS CFX for alternative reactor concepts is described and first coupled steam and helium simulations are

  11. The coupled code system TORT-TD/ATTICA3D for 3-D transient analysis of pebble-bed HTGR

    Seubert, A.; Sureda, A.; Lapins, J.; Buck, M.; Laurien, E.; Bader, J.; EnBW Kernkraft GmbH, Philippsburg

    2012-01-01

    This paper describes the time-dependent 3-D discrete-ordinates based coupled code system TORT-TD/ATTICA3D and its application to HTGR of pebble bed type. TORT-TD/ATTICA3D is represented by a single executable and adapts the so-called internal coupling approach. Three-dimensional distributions of temperatures from ATTICA3D and power density from TORT-TD are efficiently exchanged by direct memory access of array elements via interface routines. Applications of TORT-TD/ATTICA3D to three transients based on the PBMR-400 benchmark (total and partial control rod withdrawal and cold helium ingress) and the full power steady state of the HTR-10 are presented. For the partial control rod withdrawal, 3-D effects of local neutron flux redistributions are clearly identified. The results are very promising and demonstrate that the coupled code system TORT-TD/ATTICA3D may represent a key component in a future comprehensive 3-D code system for HTGR of pebble bed type. (orig.)

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

    Schuerenkraemer, M.

    1984-04-01

    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 300 0 C 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 1300 0 C. (orig.) [de

  13. Hot helium flow test facility summary report

    1980-06-01

    This report summarizes the results of a study conducted to assess the feasibility and cost of modifying an existing circulator test facility (CTF) at General Atomic Company (GA). The CTF originally was built to test the Delmarva Power and Light Co. steam-driven circulator. This circulator, as modified, could provide a source of hot, pressurized helium for high-temperature gas-cooled reactor (HTGR) and gas-cooled fast breeder reactor (GCFR) component testing. To achieve this purpose, a high-temperature impeller would be installed on the existing machine. The projected range of tests which could be conducted for the project is also presented, along with corresponding cost considerations

  14. Gas turbine modular helium reactor in cogeneration

    Leon de los Santos, G.

    2009-10-01

    This work carries out the thermal evaluation from the conversion of nuclear energy to electric power and process heat, through to implement an outline gas turbine modular helium reactor in cogeneration. Modeling and simulating with software Thermo flex of Thermo flow the performance parameters, based on a nuclear power plant constituted by an helium cooled reactor and helium gas turbine with three compression stages, two of inter cooling and one regeneration stage; more four heat recovery process, generating two pressure levels of overheat vapor, a pressure level of saturated vapor and one of hot water, with energetic characteristics to be able to give supply to a very wide gamma of industrial processes. Obtaining a relationship heat electricity of 0.52 and efficiency of net cogeneration of 54.28%, 70.2 MW net electric, 36.6 MW net thermal with 35% of condensed return to 30 C; for a supplied power by reactor of 196.7 MW; and with conditions in advanced gas turbine of 850 C and 7.06 Mpa, assembly in a shaft, inter cooling and heat recovery in cogeneration. (Author)

  15. Helium-flow measurement using ultrasonic technique

    Sondericker, J.H.

    1983-01-01

    While designing cryogenic instrumentation for the Colliding Beam Accelerator (CBA) helium-distribution system it became clear that accurate measurement of mass flow of helium which varied in temperature from room to sub-cooled conditions would be difficult. Conventional venturi flow meters full scale differential pressure signal would decrease by more than an order of magnitude during cooldown causing unacceptable error at operating temperature. At sub-cooled temperatures, helium would be pumped around cooling loops by an efficient, low head pressure circulating compressor. Additional pressure drop meant more pump work was necessary to compress the fluid resulting in a higher outlet temperature. The ideal mass flowmeter for this application was one which did not add pressure drop to the system, functioned over the entire temperature range, has high resolution and delivers accurate mass flow measurement data. Ultrasonic flow measurement techniques used successfully by the process industry, seemed to meet all the necessary requirements. An extensive search for a supplier of such a device found that none of the commercial stock flowmeters were adaptable to cryogenic service so the development of the instrument was undertaken by the CBA Cryogenic Control and Instrumentation Engineering Group at BNL

  16. Ultralow temperature helium compressor for Japan Atomic Energy Research Institute

    Asakura, Hiroshi

    1988-01-01

    Ishikawajima Harima Heavy Industries Co., Ltd. started the development of an ultralow temperature helium compressor for helium liquefaction in 1984 jointly with Japan Atomic Energy Research Institute, and has delivered the first practical machine to the Superconductive Magnet Laboratory of JAERI. For a large superconductive magnet to be used in the stable state for a fusion reactor, conventional superconductive materials (NbTi, NbTi 3 Sn, etc.) must be used, being cooled forcibly with supercritical helium. The supercritical helium which is compressed above the critical pressure of 228 kPa has a stable cooling effect since the thermal conductivity does not change due to the evaporation of liquid helium. In order to maintain the temperature of the supercritical helium below 4 K before it enters a magnet, a heat exchanger is used. The compressor that IHI has developed has the ability to reduce the vapor pressure of liquid helium from atmospheric pressure to 50.7 kPa, and can attain the temperature of 3.5 K. The specification of this single stage centrifugal compressor is: mass flow rate 25 - 64 g/s, speed 80,000 rpm, adiabatic efficiency 62 - 69 %. The structure and the performance are reported. (K.I.)

  17. State of the Art Report for a Bearing for VHTR Helium Circulator

    Lee, Jae Seon; Song, Kee Nam; Kim, Yong Wan; Lee, Won Jae

    2008-10-01

    A helium circulator in a VHTR(Very High Temperature gas-cooled Reactor) plays a core role which translates thermal energy at high temperature from a nuclear core to a steam generator. Helium as a operating coolant circulates a primary circuit in high temperature and high pressure state, and controls thermal output of a nuclear core by controlling flow rate. A helium circulator is the only rotating machinery in a VHTR, and its reliability should be guaranteed for reliable operation of a reactor and stable production of hydrogen. Generally a main helium circulator is installed on the top of a steam generator vessel, and helium is circulated only by a main helium circulator in a normal operation state. An auxiliary or shutdown circulator is installed at the bottom of a reactor vessel, and it is an auxiliary circulator for shutting down a reactor in case of refueling or accelerating cooling down in case of fast cooling. Since a rotating shaft of a helium circulator is supported by bearings, bearings are the important machine elements which determines reliability of a helium circulator and a nuclear reactor. Various types of support bearings have been developed and applied for circulator bearings since 1960s, and it is still developing for developing VHTRs. So it is necessary to review and analyze the current technical state of helium circulator support bearings to develop bearings for Koran developing VHTR helium circulator

  18. A liquid helium saver

    Avenel, O.; Der Nigohossian, G.; Roubeau, P.

    1976-01-01

    A cryostat equipped with a 'liquid helium saver' is described. A mass flow rate M of helium gas at high pressure is injected in a counter-flow heat exchanger extending from room to liquid helium temperature. After isenthalpic expansion through a calibrated flow impedance this helium gas returns via the low pressure side of the heat exchanger. The helium boil-off of the cryostat represents a mass flow rate m, which provides additional precooling of the incoming helium gas. Two operating regimes appear possible giving nearly the same efficiency: (1) high pressure (20 to 25 atm) and minimum flow (M . L/W approximately = 1.5) which would be used in an open circuit with helium taken from a high pressure cylinder; and (2) low pressure (approximately = 3 atm), high flow (M . L/W > 10) which would be used in a closed circuit with a rubber diaphragm pumping-compressing unit; both provide a minimum theoretical boil-off factor of about 8%. Experimental results are reported. (U.K.)

  19. Innovative technologies for Faraday shield cooling

    Rosenfeld, J.H.; Lindemuth, J.E.; North, M.T.; Goulding, R.H.

    1995-01-01

    Alternative advanced technologies are being evaluated for use in cooling the Faraday shields used for protection of ion cyclotron range of frequencies (ICR) antennae in Tokamaks. Two approaches currently under evaluation include heat pipe cooling and gas cooling. A Monel/water heat pipe cooled Faraday shield has been successfully demonstrated. Heat pipe cooling offers the advantage of reducing the amount of water discharged into the Tokamak in the event of a tube weld failure. The device was recently tested on an antenna at Oak Ridge National Laboratory. The heat pipe design uses inclined water heat pipes with warm water condensers located outside of the plasma chamber. This approach can passively remove absorbed heat fluxes in excess of 200 W/cm 2 ;. Helium-cooled Faraday shields are also being evaluated. This approach offers the advantage of no liquid discharge into the Tokamak in the event of a tube failure. Innovative internal cooling structures based on porous metal cooling are being used to develop a helium-cooled Faraday shield structure. This approach can dissipate the high heat fluxes typical of Faraday shield applications while minimizing the required helium blower power. Preliminary analysis shows that nominal helium flow and pressure drop can sufficiently cool a Faraday shield in typical applications. Plans are in progress to fabricate and test prototype hardware based on this approach

  20. Parametric study for high conversion pebble bed reactor

    Teuchert, E.; Ruetten, H. J.

    1975-06-15

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

  1. Uraniferous quartz-pebble conglomerates in South Africa

    von Backstroem, J.W.

    1981-01-01

    The purpose of this paper is to give a short background statement summarizing data on the Dominion Reef Group, the Witwatersrand Supergroup, and the Ventersdorp Contact Reef, with particular reference to the close relationship of gold and uranium with sedimentary features as well as the mineralization, conditions of deposition, and the nature of the quartz-pebble conglomerates

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

    Bende, E.E.

    1997-01-01

    The pebble bed type High Temperature Reactor (HTR) has been studied as a uranium-free burner of reactor grade plutonium. In a parametric study, the plutonium loading per pebble as well as the type and size of the coated particles (CPs) have been varied to determine the plutonium consumption, the final plutonium burnup, the k ∞ and the temperature coefficients as a function of burnup. The plutonium loading per pebble is bounded between 1 and 3 gr Pu per pebble. The upper limit is imposed by the maximal allowable fast fluence for the CPs. A higher plutonium loading requires a longer irradiation time to reach a desired burnup, so that the CPs are exposed to a higher fast fluence. The lower limit is determined by the temperature coefficients, which become less negative with increasing moderator-actinide ratio. A burnup of about 600 MWd/kgHM can be reached. With the HTR's high efficiency of 40%, a plutonium supply of 1520 kg/GW e a is achieved. The discharges of plutonium and minor actinides are then 450 and 110 kg/GW e a, respectively. (author)

  3. Performance Evaluation of a Pebble Bed Solar Crop Dryer ...

    Nigerian Journal of Technology ... The solar crop dryer consists of an imbedded pebble bed solar heat storage unit/solar collector ... The crop-drying chamber is made of drying trays of wire gauze while the roof is made of transparent glazing.

  4. Spectroscopy of antiproton helium atoms

    Hayano, Ryugo

    2005-01-01

    Antiproton helium atom is three-body system consisting of an antiproton, electrons and a helium nucleus (denoted by the chemical symbol, p-bar H + ). The authors produced abundant atoms of p-bar 4 He + , and p-bar 3 He + in a cooled He gas target chamber stopping the p-bar beam decelerated to approximately 100 keV in the Antiproton Decelerator at CERN. A precision laser spectroscopy on the atomic transitions in the p-bar 4 He + , and in p-bar 3 He + was performed. Principle of laser spectroscopy and various modifications of the system to eliminate factors affecting the accuracy of the experiment were described. Deduced mass ratio of antiproton and proton, (|m p -bar - m p |)/m p reached to the accuracy of 10 ppb (10 -8 ) as of 2002, as adopted in the recent article of the Particle Data Group by P.J. Mohr and B.N. Taylor. This value is the highest precise data for the CPT invariance in baryon. In future, antihydrogen atoms will be produced in the same facility, and will provide far accurate value of antiproton mass thus enabling a better confirmation of CPT theorem in baryon. (T. Tamura)

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

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

    2013-01-01

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

  6. Fabrication of Li2TiO3 pebbles by a freeze drying process

    Lee, Sang-Jin; Park, Yi-Hyun; Yu, Min-Woo

    2013-01-01

    Li 2 TiO 3 pebbles were successfully fabricated by using a freeze drying process. The Li 2 TiO 3 slurry was prepared using a commercial powder of particle size 0.5–1.5 μm and the pebble pre-form was prepared by dropping the slurry into liquid nitrogen through a syringe needle. The droplets were rapidly frozen, changing their morphology to spherical pebbles. The frozen pebbles were dried at −10 °C in vacuum. To make crack-free pebbles, some glycerin was employed in the slurry, and long drying time and a low vacuum condition were applied in the freeze drying process. In the process, the solid content in the slurry influenced the spheroidicity of the pebble green body. The dried pebbles were sintered at 1200 °C in an air atmosphere. The sintered pebbles showed almost 40% shrinkage. The sintered pebbles revealed a porous microstructure with a uniform pore distribution and the sintered pebbles were crushed under an average load of 50 N in a compressive strength test. In the present study, a freeze drying process for fabrication of spherical Li 2 TiO 3 pebbles is introduced. The processing parameters, such as solid content in the slurry and the conditions of freeze drying and sintering, are also examined

  7. Influence of hydrogen addition to a sweep gas on tritium behavior in a blanket module containing Li{sub 2}TiO{sub 3} pebbles

    Katayama, K., E-mail: kadzu@nucl.kyushu-u.ac.jp [Department of Advanced Energy Engineering Science, Kyushu University 6-1, Kasugakoen, Kasuga-shi, Fukuoka 816-8580 (Japan); Someya, Y.; Tobita, K. [National Institutes for Quantum and radiological Science and Technology, 2-166 Omotedate, Obuchi, Rokkasho-mura, Kamikita-gun, Aomori 039-3212 (Japan); Fukada, S. [Department of Advanced Energy Engineering Science, Kyushu University 6-1, Kasugakoen, Kasuga-shi, Fukuoka 816-8580 (Japan); Hatano, Y. [Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555 (Japan); Chikada, T. [Department of Chemistry, Graduate school of Science, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka, 422-8529 (Japan)

    2016-12-15

    Highlights: • Mass balance equations of H{sub 2}, H{sub 2}O, T{sub 2} and T{sub 2}O in a Li{sub 2}TiO{sub 3} pebble bed were numerically calculated. • In the temperature rising process, the pebbles were exposed to water vapor of relatively high concentration. • Tritium permeation rate to cooling water reduced with increasing hydrogen concentration in the sweep gas. • Tritium inventory in the grain bulk and the grain surface occupied 99.6% of total inventory. - Abstract: Hydrogen addition to a sweep gas of a solid breeder blanket module has been proposed to enhance tritium recovery from the surface of the breeders. However, the influence of hydrogen addition on the bred tritium behavior is not understood completely. Tritium behavior in the simplified blanket module of Li{sub 2}TiO{sub 3} pebbles was numerically calculated considering diffusion in the grain bulk, surface reactions on the grain s