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Sample records for lead-bismuth cooled fast

  1. Performance Comparison of Metallic, Actinide Burning Fuel in Lead-Bismuth and Sodium Cooled Fast Reactors

    SciTech Connect

    Weaver, Kevan Dean; Herring, James Stephen; Mac Donald, Philip Elsworth

    2001-04-01

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

  2. Design study of lead bismuth cooled fast reactors and capability of natural circulation

    SciTech Connect

    Oktamuliani, Sri Su’ud, Zaki

    2015-09-30

    A preliminary study designs SPINNOR (Small Power Reactor, Indonesia, No On-Site Refueling) liquid metal Pb-Bi cooled fast reactors, fuel (U, Pu)N, 150 MWth have been performed. Neutronic calculation uses SRAC which is designed cylindrical core 2D (R-Z) 90 × 135 cm, on the core fuel composed of heterogeneous with percentage difference of PuN 10, 12, 13% and the result of calculation is effective neutron multiplication 1.0488. Power density distribution of the output SRAC is generated for thermal hydraulic calculation using Delphi based on Pascal language that have been developed. The research designed a reactor that is capable of natural circulation at inlet temperature 300 °C with variation of total mass flow rate. Total mass flow rate affect pressure drop and temperature outlet of the reactor core. The greater the total mass flow rate, the smaller the outlet temperature, but increase the pressure drop so that the chimney needed more higher to achieve natural circulation or condition of the system does not require a pump. Optimization of the total mass flow rate produces optimal reactor design on the total mass flow rate of 5000 kg/s with outlet temperature 524,843 °C but require a chimney of 6,69 meters.

  3. Design study of lead bismuth cooled fast reactors and capability of natural circulation

    NASA Astrophysics Data System (ADS)

    Oktamuliani, Sri; Su'ud, Zaki

    2015-09-01

    A preliminary study designs SPINNOR (Small Power Reactor, Indonesia, No On-Site Refueling) liquid metal Pb-Bi cooled fast reactors, fuel (U, Pu)N, 150 MWth have been performed. Neutronic calculation uses SRAC which is designed cylindrical core 2D (R-Z) 90 × 135 cm, on the core fuel composed of heterogeneous with percentage difference of PuN 10, 12, 13% and the result of calculation is effective neutron multiplication 1.0488. Power density distribution of the output SRAC is generated for thermal hydraulic calculation using Delphi based on Pascal language that have been developed. The research designed a reactor that is capable of natural circulation at inlet temperature 300 °C with variation of total mass flow rate. Total mass flow rate affect pressure drop and temperature outlet of the reactor core. The greater the total mass flow rate, the smaller the outlet temperature, but increase the pressure drop so that the chimney needed more higher to achieve natural circulation or condition of the system does not require a pump. Optimization of the total mass flow rate produces optimal reactor design on the total mass flow rate of 5000 kg/s with outlet temperature 524,843 °C but require a chimney of 6,69 meters.

  4. Design of alumina forming FeCrAl steels for lead or lead-bismuth cooled fast reactors

    NASA Astrophysics Data System (ADS)

    Lim, Jun; Hwang, Il Soon; Kim, Ji Hyun

    2013-10-01

    Iron-chromium-aluminum alloys containing 15-20 wt.% Cr and 4-6 wt.% Al have shown excellent corrosion resistance in the temperature range up to 600 °C or higher in liquid lead and lead-bismuth eutectic environments by the formation of protective Al2O3 layers. However, the higher Cr and Al concentrations in ferritic alloys could be problematic because of severe embrittlement in the manufacturing process as well as in service, caused by the formation of brittle phases. For this reason, efforts worldwide have so far mainly focused on the development of aluminizing surface treatments. However, aluminizing surface treatments have major disadvantages of cost, processing difficulties and reliability issues. In this study, a new FeCrAl alloy is proposed for structural materials in lead and lead-bismuth cooled nuclear applications. The alloy design relied on corrosion experiments in high temperature lead and lead-bismuth eutectic environments and computational thermodynamic calculations using the commercial software, JMatPro. The design of new alloys has focused on the optimization of Cr and Al levels for the formation of an external Al2O3 layer which can provide excellent oxidation and corrosion resistance in liquid lead alloys in the temperature range 300-600 °C while still retaining workable mechanical properties.

  5. Preliminary study on nano- and micro-composite sol-gel based alumina coatings on structural components of lead-bismuth eutectic cooled fast breeder reactors

    NASA Astrophysics Data System (ADS)

    Dou, Peng; Kasada, Ryuta

    2011-02-01

    In order to protect the structural components of lead-bismuth eutectic cooled fast breeder reactors from liquid metal corrosion, Al 2O 3 nano- and micro-composite coatings were developed using an improved sol-gel process, which includes dipping specimens in a sol-gel solution dispersed with fine α-Al 2O 3 powders prepared by mechanical milling. Accelerated corrosion tests were conducted on coated specimens in liquid lead-bismuth eutectic at 500 °C under dynamic conditions. Scanning electron microscopy (SEM) and X-ray diffraction (XRD) analyses revealed that the coatings are composed of α-Al 2O 3 and they are about 10 μm thick. After the corrosion tests, no spallation occurred on the coatings, and neither Pb nor Bi penetrated into the coatings, which indicates that the coatings possess an enhanced dynamic LBE corrosion resistance to lead-bismuth eutectic corrosion. The nano-structured composite particles integrated into the coatings play an important role in achieving such superior lead-bismuth eutectic corrosion resistance.

  6. Sol-gel composite coatings as anti-corrosion barrier for structural materials of lead-bismuth eutectic cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Kasada, Ryuta; Dou, Peng

    2013-09-01

    In order to protect the structural components of lead-bismuth eutectic (LBE) cooled fast breeder reactors (FBRs) from liquid metal corrosion, advanced aluminum-yttrium nano- and micro-composite coatings were developed using an improved sol-gel process, which includes dipping specimens in a Y-added sol-gel solution dispersed with ultrafine α-Al2O3 powders prepared by mechanical milling. Scanning electron microscopy (SEM) and field emission electron probe microprobe analyzer (FE-EPMA) analyses revealed that the coatings are composed of alumina with high density. Accelerated corrosion tests were conducted on coated specimens in liquid LBE at 650 °C under dynamic conditions. After the corrosion tests, no cracking, spallation, erosion and liquid metal (e.g., lead) penetration occurred to the coatings, indicating that the coatings possess an enhanced dynamic LBE corrosion resistance. The superior LBE corrosion resistance is due to the presence of the nano-structured composite particles integrated into the coatings and the addition of trace amount of yttrium. Severe erosion and penetration of liquid Pb occurred to the Al2O3 nano- and micro-composite coatings. After the corrosion tests, no cracking, spallation, erosion and liquid metal (e.g., lead) penetration occurred to the newly-developed aluminum-yttrium nano- and micro-composite coatings, indicating that the coatings possess an enhanced dynamic LBE corrosion resistance. Therefore we can conclude that the coatings possess an enhanced dynamic LBE corrosion resistance under the experimental conditions chosen here. It is a way to protect the structural materials of LBE cooled FBRs from liquid metal corrosion. The much improved corrosion resistance of aluminum-yttrium nano- and micro-composite coatings, relative to Al2O3 nano- and micro-composite coatings, is due to the much higher density and the significantly superior high temperature strength resulting from using of finer Al2O3 seeding particles and adding trace

  7. Design of an Actinide Burning, Lead-Bismuth Cooled Reactor That Produces Low Cost Electricity

    SciTech Connect

    C. Davis; S. Herring; P. MacDonald; K. McCarthy; V. Shah; K. Weaver; J. Buongiorno; R. Ballinger; K. Doyoung; M. Driscoll; P. Hejzler; M. Kazimi; N. Todreas

    1999-07-01

    The purpose of this project is to investigate the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. The choice of lead-bismuth for the reactor coolant is an actinide burning fast reactor offers enhanced safety and reliability. The advantages of lead-bismuth over sodium as a coolant are related to the following material characteristics: chemical inertness with air and water; higher atomic number; lower vapor pressure at operating temperatures; and higher boiling temperature. Given the status of the field, it was agreed that the focus of this investigation in the first two years will be on the assessment of approaches to optimize core and plant arrangements in order to provide maximum safety and economic potential in this type of reactor.

  8. One-group fission cross sections for plutonium and minor actinides inserted in calculated neutron spectra of fast reactor cooled with lead-208 or lead-bismuth eutectic

    SciTech Connect

    Khorasanov, G. L.; Blokhin, A. I.

    2012-07-01

    The paper is dedicated to one-group fission cross sections of Pu and MA in LFRs spectra with the aim to increase these values by choosing a coolant which hardens neutron spectra. It is shown that replacement of coolant from Pb-Bi with Pb-208 in the fast reactor RBEC-M, designed in Russia, leads to increasing the core mean neutron energy. As concerns fuel Pu isotopes, their one-group fission cross sections become slightly changed, while more dramatically Am-241 one-group fission cross section is changed. Another situation occurs in the lateral blanket containing small quantities of minor actinides. It is shown that as a result of lateral blanket mean neutron energy hardening the one-group fission cross sections of Np-237, Am-241 and Am-243 increases up to 8-11%. This result allows reducing the time of minor actinides burning in FRs. (authors)

  9. Studies of Polonium Removal from Molten Lead-Bismuth for Lead-Alloy-Cooled Reactor Applications

    SciTech Connect

    Jacopo Buongiorno; Ken Czerwinski; Eric Loewen; Chris Larson

    2004-09-01

    The isotope 210Po is the main product of neutron activation in fast reactors cooled by molten lead-bismuth eutectic (LBE). The isotope 210Po is a pure alpha emitter with a half-life of 138.38 days. For typical values of the neutron flux the 210Po concentration in the coolant can reach 1-10 Ci/kg. While exposure of plant personnel to Po is prevented under normal operating conditions because the primary system is sealed, Po does pose a radiological hazard during maintenance activities for which access to submerged structures is required as well as during accidents resulting in breach of the primary-system barrier. Obviously, continuous removal of Po from the LBE reduces this hazard. Therefore, it is important to understand the mechanisms by which Po is formed in and released from the LBE. We summarize research performed at the Idaho National Engineering and Environmental Laboratory and the Massachusetts Institute of Technology to investigate the basic chemistry of four mechanisms of Po release, which could serve as the basis for a coolant cleanup system in LBE-cooled reactors. The mechanisms explored are lead polonide evaporation, formation of polonium hydride, rare-earth filtering, and alkaline extraction. For the key chemical species involved expressions are given for useful quantities such as formation energy, release, and deposition rates. It is concluded that the most promising removal mechanism is alkaline extraction, although a more systematic investigation of this mechanism is needed.

  10. Studies of Polonium Removal from Molten Lead-Bismuth for Lead-Alloy-Cooled Reactor Applications

    SciTech Connect

    Buongiorno, Jacopo; Loewen, Eric P.; Czerwinski, Kenneth; Larson, Christopher

    2004-09-15

    The isotope {sup 210}Po is the main product of neutron activation in fast reactors cooled by molten lead-bismuth eutectic (LBE). The isotope {sup 210}Po is a pure alpha emitter with a half-life of 138.38 days. For typical values of the neutron flux the {sup 210}Po concentration in the coolant can reach 1-10 Ci/kg. While exposure of plant personnel to Po is prevented under normal operating conditions because the primary system is sealed, Po does pose a radiological hazard during maintenance activities for which access to submerged structures is required as well as during accidents resulting in breach of the primary-system barrier. Obviously, continuous removal of Po from the LBE reduces this hazard. Therefore, it is important to understand the mechanisms by which Po is formed in and released from the LBE. We summarize research performed at the Idaho National Engineering and Environmental Laboratory and the Massachusetts Institute of Technology to investigate the basic chemistry of four mechanisms of Po release, which could serve as the basis for a coolant cleanup system in LBE-cooled reactors. The mechanisms explored are lead polonide evaporation, formation of polonium hydride, rare-earth filtering, and alkaline extraction. For the key chemical species involved expressions are given for useful quantities such as formation energy, release, and deposition rates. It is concluded that the most promising removal mechanism is alkaline extraction, although a more systematic investigation of this mechanism is needed.

  11. System Analysis for Decay Heat Removal in Lead-Bismuth-Cooled Natural-Circulation Reactors

    SciTech Connect

    Sakai, Takaaki; Enuma, Yasuhiro; Iwasaki, Takashi

    2004-03-15

    Decay heat removal analyses for lead-bismuth-cooled natural-circulation reactors are described in this paper. A combined multidimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural-circulation reactors. For the preliminary study, transient analysis has been performed for a 300-MW(thermal) lead-bismuth-cooled reactor designed by Argonne National Laboratory. In addition, decay heat removal characteristics of a 400-MW(electric) lead-bismuth-cooled natural-circulation reactor designed by the Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. The primary reactor auxiliary cooling system (PRACS) is prepared for the JNC concept to get sufficient heat removal capacity. During 2000 s after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 deg. C because the buoyancy force in a primary circulation path is temporarily reduced. However, the natural circulation is recovered by the PRACS system, and the outlet temperature decreases successfully.

  12. System Analysis for Decay Heat Removal in Lead-Bismuth Cooled Natural Circulated Reactors

    SciTech Connect

    Takaaki Sakai; Yasuhiro Enuma; Takashi Iwasaki; Kazuhiro Ohyama

    2002-07-01

    Decay heat removal analyses for lead-bismuth cooled natural circulation reactors are described in this paper. A combined multi-dimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural circulation reactors. For the preliminary study, transient analysis has been performed for a 100 MWe lead-bismuth-cooled reactor designed by Argonne National Laboratory (ANL). In addition, decay heat removal characteristics of a 400 MWe lead-bismuth-cooled natural circulation reactor designed by Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. PRACS (Primary Reactor Auxiliary Cooling System) is prepared for the JNC's concept to get sufficient heat removal capacity. During 2000 sec after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 Centigrade, because the buoyancy force in a primary circulation path is temporary reduced. However, the natural circulation is recovered by the PRACS system and the out let temperature decreases successfully. (authors)

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

    PubMed

    De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

    2016-01-01

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation. PMID:27110782

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

    PubMed Central

    De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

    2016-01-01

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation. PMID:27110782

  15. Design of an Actinide Burning, Lead or Lead-Bismuth Cooled Reactor that Produces Low Cost Electricity FY-01 Annual Report, October 2001

    SciTech Connect

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Herring, James Stephen; Loewen, Eric Paul; Smolik, Galen Richard; Weaver, Kevan Dean; Todreas, N.

    2001-10-01

    The purpose of this collaborative Idaho National Engineering and Environmental Laboratory (INEEL) and Massachusetts Institute of Technology (MIT) Laboratory Directed Research and Development (LDRD) project is to investigate the suitability of lead or lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. Work has been accomplished in four major areas of research: core neutronic design, plant engineering, material compatibility studies, and coolant activation. The publications derived from work on this project (since project inception) are listed in Appendix A.

  16. Design of an Actinide Burning, Lead or Lead-Bismuth Cooled Reactor That Produces Low Cost Electricty - FY-02 Annual Report

    SciTech Connect

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo

    2002-10-01

    The purpose of this collaborative Idaho National Engineering and Environmental Laboratory (INEEL) and Massachusetts Institute of Technology (MIT) Laboratory Directed Research and Development (LDRD) project is to investigate the suitability of lead or lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. Work has been accomplished in four major areas of research: core neutronic design, plant engineering, material compatibility studies, and coolant activation. The publications derived from work on this project (since project inception) are listed in Appendix A. This is the third in a series of Annual Reports for this project, the others are also listed in Appendix A as FY-00 and FY-01 Annual Reports.

  17. Overview of the Use of ATHENA for Thermal-Hydraulic Analysis of Systems with Lead-Bismuth Coolant

    SciTech Connect

    Davis, Cliff Bybee; Shieh, Arthur Shan Luk

    2000-04-01

    The INEEL and MIT are investigating the suitability of lead-bismuth cooled fast reactor for producing low-cost electricity as well as for actinide burning. This paper is concerned with the general area of thermal-hydraulics of lead-bismuth cooled reactors. The ATHENA code is being used in the thermal-hydraulic design and analysis of lead-bismuth cooled reactors. The ATHENA code was reviewed to determine its applicability for simulating lead-bismuth cooled reactors. Two modifications were made to the code as a result of this review. Specifically, a correlation to represent heat transfer from rod bundles to a liquid metal and a void correlation based on data taken in a mixture of lead-bismuth and steam were added the code. The paper also summarizes the analytical work that is being performed with the code and plans for future analytical work.

  18. Overview of the use of ATHENA for thermal-hydraulic analysis of systems with lead-bismuth coolant

    SciTech Connect

    C. B. Davis; A. S. Shieh

    2000-04-02

    The INEEL and MIT are investigating the suitability of lead-bismuth cooled fast reactor for producing low-cost electricity as well as for actinide burning. This paper is concerned with the general area of thermal-hydraulics of lead-bismuth cooled reactors. The ATHENA code is being used in the thermal-hydraulic design and analysis of lead-bismuth cooled reactors. The ATHENA code was reviewed to determine its applicability for simulating lead-bismuth cooled reactors. Two modifications were made to the code as a result of this review. Specifically, a correlation to represent heat transfer from rod bundles to a liquid metal and a void correlation based on data taken in a mixture of lead-bismuth and steam were added the code. The paper also summarizes the analytical work that is being performed with the code and plans for future analytical work.

  19. Transient Thermo-Hydraulic Analysis of the Windowless Target System for the Lead Bismuth Eutectic Cooled Accelerator Driven System

    SciTech Connect

    Bianchi, Fosco; Ferri, Roberta; Moreau, Vincent

    2006-07-01

    The target system, whose function is to supply an external neutron source to the ADS sub-critical core to sustain the neutron chain reaction, is the most critical part of an ADS being subject to severe thermo-mechanical loading and material damage due to accelerator protons and fission neutrons. A windowless option was chosen as reference configuration for the target system of the LBE-cooled ADS within the European PDS-XADS project in order to reduce the material damage and to increase its life. This document deals with the thermo-hydraulic results of the calculations performed with STAR-CD and RELAP5 codes for studying the behaviour of the windowless target system during off-normal operating conditions. It also reports a description of modifications properly implemented in the codes needed for this analysis. The windowless target system shows a satisfactory thermo-hydraulic behaviour for the analysed accidents, except for the loss of both pumps without proton beam shut-off and the beam trips lasting more than one second. (authors)

  20. Ultrasound in lead-bismuth eutectic

    SciTech Connect

    Dierckx, M.; Van Dyck, D.

    2011-07-01

    The Belgian Nuclear Research Centre (SCK.CEN) is in the process of designing MYRRHA, a new multi-purpose irradiation facility to replace the ageing BR2. MYRRHA is a fast spectrum reactor cooled with lead-bismuth eutectic (LBE). As liquid metal is opaque to visual light, ultrasonic measurement techniques are selected to fulfill essential tasks that, according to our assessment, will be demanded by licensing authorities, in particular: fuel assembly identification and localization of a lost fuel assembly. To that end, a considerable research effort at SCK.CEN is devoted to study ultrasonic propagation in LBE. As ultrasonic experiments in LBE are elaborate and expensive to set up, we are particularly interested in to what extent experiments in water can be extrapolated to LBE - one of the main focuses of this article. We describe and present results of a first experiment with this goal which shows that the signal to noise ratio is better in LBE and that we even see small diffuse reflections up to 40 deg. off normal. On the other hand, we do not see internal reflections in stainless steel objects in LBE which we do in water. Therefore, we conclude that experiments in water can be used to validate algorithms for LBE on the condition that they do not rely on internal reflections. We also present solutions to tackle the essential tasks: fuel assembly identification and lost object localization. The requirements for the ultrasonic equipment implementing these solutions are also discussed. (authors)

  1. Conceptual Design of Passive Safety System for Lead-Bismuth Cooled Fast Reactor

    NASA Astrophysics Data System (ADS)

    Abdullah, A. G.; Nandiyanto, A. B. D.

    2016-04-01

    This paper presents the results of the conceptual design of passive safety systems for reactor power 225 MWth using Pb-Bi coolant. Main purpose of this research is to design of heat removal system from the reactor wall. The heat from the reactor wall is removed by RVACS system using the natural circulation from the atmosphere around the reactor at steady state. The calculation is performed numerically using Newton-Raphson method. The analysis involves the heat transfer systems in a radiation, conduction and natural convection. Heat transfer calculations is performed on the elements of the reactor vessel, outer wall of guard vessel and the separator plate. The simulation results conclude that the conceptual design is able to remove heat 1.33% to 4.67% from the thermal reactor power. It’s can be hypothesized if the reactor had an accident, the system can still overcome the heat due to decay.

  2. Design study of lead- and lead-bismuth-cooled small long-life nuclear power reactors using metallic and nitride fuel

    SciTech Connect

    Sekimoto, Hiroshi; Su`ud, Zaki

    1995-03-01

    A conceptual design study of small long-life nuclear power reactors used for a remote or isolated area has been performed. Lead as well as lead-bismuth is employed as the coolant, and both metallic and nitride fuels are investigated. There are some severe requirements on these reactors for operability, maintainability, safety, and proliferation resistance. Some important characteristics of the proposed designs [150 MW (thermal)] are the following: transportability between reactor factory and operation site; capability of long-life operation (12 yr) without refueling or fuel shuffling while maintaining burnup reactivity swing less than 0.1% {Delta}k; negative total core coolant void coefficient of reactivity over all the burnup period; omission of intermediate heat exchanger; and a relatively large contribution of natural circulation.

  3. Lead-bismuth eutectic technology for Hyperion reactor

    NASA Astrophysics Data System (ADS)

    Zhang, J.; Kapernick, R. J.; McClure, P. R.; Trapp, T. J.

    2013-10-01

    A small lead-bismuth eutectic-cooled reactor concept (referred to as the Hyperion reactor concept) is being studied at Los Alamos National Laboratory and Hyperion Power Generation. In this report, a critical assessment of the lead-bismuth eutectic technology for Hyperion reactor is presented based on currently available knowledge. Included are: material compatibility, oxygen control, thermal hydraulics, polonium control. The key advances in the technology and their applications to Hyperion reactor design are analyzed. Also, the near future studies in main areas of the technology are recommended for meeting the design requirements.

  4. Numerical Analysis of Lead-Bismuth-Water Direct Contact Boiling Heat Transfer

    NASA Astrophysics Data System (ADS)

    Yamada, Yumi; Takahashi, Minoru

    Direct contact boiling heat transfer of sub-cooled water with lead-bismuth eutectic (Pb-Bi) was investigated for the evaluation of the performance of steam generation in direct contact of feed water with primary Pb-Bi coolant in upper plenum above the core in Pb-Bi-cooled direct contact boiling water fast reactor. An analytical two-fluid model was developed to estimate the heat transfer numerically. Numerical results were compared with experimental ones for verification of the model. The overall volumetric heat transfer coefficient was calculated from heat exchange rate in the chimney. It was confirmed that the calculated results agreed well with the experimental result.

  5. Design of Material Strength Test in Lead-Bismuth Flow

    SciTech Connect

    Masatoshi Kondo; Minoru Takahashi; Koji Hata

    2002-07-01

    Liquid lead and lead-bismuth have drawn the attention as one of the candidate coolants of the fast breeder reactors (FBRs), and the accelerator driven transmutation systems (ADSs). In order to use the coolant to the systems, the physical and chemical characteristics of the heavy metals are necessary. This plan has been proposed for the strength test of materials in the liquid metal surroundings. The lead-bismuth circulation loop with the strength test has been designed, and the strength test of candidate materials has been planned. (authors)

  6. Lead-Bismuth Activities at the Karlsruhe Lead Laboratory KALLA

    SciTech Connect

    Knebel, Joachim U.; Muller, Georg; Konys, Jurgen

    2002-07-01

    At Forschungszentrum Karlsruhe (FZK) the characteristics of an accelerator-driven subcritical system (ADS) are evaluated, mainly with respect to the potential of transmutation of minor actinides and long-lived fission products, to the feasibility and to safety aspects. All experimental activities, which are related to lead-bismuth as cooling fluid and spallation material, are performed in the Karlsruhe Lead Laboratory KALLA. This article gives an overview on KALLA, which has three stagnant experiments and three loop experiments. The stagnant experiments are concentrating on corrosion mechanisms, surface treatment, oxygen sensor development, and oxygen control system (OCS), the loop experiments are concentrating on thermalhydraulic measurement techniques, ADS-relevant component testing, and corrosion investigations in flowing lead-bismuth. A fourth loop experiment is planned to investigate the integral heat removal from a 4 MW spallation target for normal and decay heat removal conditions. Among others, latest results are presented of: characteristics of oxygen sensors in flowing liquid Pb-Bi, the oxygen control system (OCS) operating on a loop system, an ultrasonic flow meter applied to lead-bismuth at 400 deg. C. In addition, results are given on the improvement of the corrosion resistivity of steels in flowing lead-bismuth, using a special temperature treatment method (electron beam facility GESA) and alloying aluminium in the surface layer. (authors)

  7. The experience in handling of lead-bismuth coolant contaminated by Polonium-210

    SciTech Connect

    Pankratov, D.V.; Gromov, B.F.; Solodjankin, M.A.

    1993-12-31

    During exploitation of lead-bismuth cooled reactors a wide experience in handling of radioactive coolant containing polonium has been gained. By 1990 total time of this reactor operation has reached approximately 60 reactor years.

  8. Comparison of Lead-Bismuth and Lead as Coolants for Accelerator Driven Systems

    SciTech Connect

    Bianchi, F.; Mattioda, F.; Meloni, P.

    2002-07-01

    In the framework of the Italian research program TRASCO (TRAsmutazione SCOrie, namely transmutation of radioactive wastes) and of the European research program PDS-XADS (Preliminary Design Study on an eXperimental Accelerator Driven System) the feasibility and operability of gas or liquid metal cooled accelerator driven system prototypes are currently under investigation. Initially the attention of the thermal-hydraulics group of ENEA research centre in Bologna has been focussed toward a lead-bismuth cooled subcritical system under natural or enhanced natural circulation according to the prototype design proposed. The interest in using lead as a coolant, which is characterized by a higher melting point, is explained by the need to increase the plant efficiency for the economic competitiveness, though the higher temperatures pose some technological problems. Moreover, the amount of activation products should result significantly lower. Of course the results obtained and the experience gained analysing the dynamical behaviour of the lead-bismuth cooled system cannot be directly transferred to lead cooled systems. This paper aims at presenting a preliminary comparison of lead-bismuth and lead in a simplified liquid metal cooled subcritical system, mainly from the thermal-hydraulics and system dynamics points of view. By means of the modified RELAP5 version, the dynamical behavior of a lead-bismuth or lead cooled system, which is intended to be a quite accurate representation of the Italian accelerator driven prototype XADS, has been studied. Although a more exhaustive comparison should take into account the necessarily different structural characteristics of lead-bismuth and lead cooled systems, the neutronic feedback on reactor power and also the slightly different neutronic properties of lead-bismuth and lead, the purely thermal-hydraulic analysis presented in this paper has shown that the dynamical behaviour of the XADS does not differ noticeable when lead is used

  9. Alkaline extraction of polonium from liquid lead bismuth eutectic

    NASA Astrophysics Data System (ADS)

    Heinitz, S.; Neuhausen, J.; Schumann, D.

    2011-07-01

    The production of highly radiotoxic polonium isotopes poses serious safety concerns for the development of future nuclear systems cooled by lead bismuth eutectic (LBE). In this paper it is shown that polonium can be extracted efficiently from LBE using a mixture of alkaline metal hydroxides (NaOH + KOH) in a temperature range between 180 and 350 °C. The extraction ratio was analyzed for different temperatures, gas blankets and phase ratios. A strong dependence of the extraction performance on the redox properties of the cover gas was found. While hydrogen facilitates the removal of polonium, oxygen has a negative influence on the extraction. These findings open new possibilities to back up the safety of future LBE based nuclear facilities.

  10. Lead-bismuth eutectic as advanced reactor collant : operational experience

    SciTech Connect

    Woloshun, K. A.; Watts, V.; Li, N.

    2004-01-01

    Some proposed advanced reactor concepts would be cooled by lead or lead-bismuth eutectic (LBE). An LBE test loop was designed and built at Los Alamos to develop the engineering and materials technology necessary to successfully implement LBE as a coolant (Fig. 1). Operational since December 2001, this test loop has been used to develop and demonstrate safe operation, oxygen concentration and metal corrosion control, instrumentation, thermal-hydraulic performance of heat exchangers and recuperators, and free convection and forced pumping. This paper discusses the technology development and lessons learned from the operation of this facility. A LBE test loop has been operational since December 2001. Using procedures, training, and engineering controls, this loop has operated without an accident. Continuous improvements in operation procedures and instrumentation over these years have resulted in a facility of high reliability, providing the groundwork for the use of LBE as a reactor coolant for temperatures up to 550 C.

  11. Corrosion behavior of cold-worked austenitic stainless steels in liquid lead-bismuth eutectic

    NASA Astrophysics Data System (ADS)

    Kurata, Yuji

    2014-05-01

    The effect of cold working on the corrosion behavior of austenitic stainless steels in liquid lead-bismuth eutectic (LBE) was studied to develop accelerator-driven systems for the transmutation of long-lived radioactive wastes and lead-bismuth cooled fast reactors. Corrosion tests on solution-treated, 20% cold-worked and 50% cold-worked 316SS and JPCA (15Cr-15Ni-Ti) were conducted in oxygen-controlled LBE. Slight ferritization caused by Ni dissolution and Pb-Bi penetration were observed for all specimens in the corrosion test conducted at 500 °C for 1000 h in liquid LBE with an intermediate oxygen concentration (1.4 × 10-7 wt.%). In the corrosion test performed at 550 °C for 1000 h in liquid LBE with a low oxygen concentration (4.2 × 10-9 wt.%), the depth of the ferritization of 316SS and JPCA increased with the extent of cold working. Only oxidation was observed in the corrosion test that was performed at 550 °C for 1000 h in liquid LBE with a high oxygen concentration (approximately 10-5 wt.%). Cold working accelerated the formation of the double layer oxide and increased the thickness of the oxide layer slightly. In contrast, the ferritization accompanied by Pb-Bi penetration was widely observed with oxidation for all specimens corrosion tested at 550 °C for 3000 h under the high-oxygen condition. Cold working increased the depth of the ferritization of 316SS and JPCA. It is considered that cold working accelerated the ferritization and Pb-Bi penetration through the enhanced dissolution of Ni into LBE due to an increase in the dislocation density under conditions in which the protective oxide layer was not formed in liquid LBE.

  12. Small LBE-Cooled Fast Reactor for Expanding Market

    SciTech Connect

    Hiroshi Sekimoto; Shinichi Makino; Kunihiko Nakamura; Yoshio Kamishima; Takashi Kawakita

    2002-07-01

    A long-life safe simple small portable proliferation-resistant reactor is expected to solve many problems associating future energy and globally environmental problems. From discussions on mainly neutronics and safety points it has been shown that the heavy liquid metal cooled fast reactor is the best candidate to satisfy the above requirements. A lead-bismuth-eutectic (LBE) cooled fast reactor LSPR (LBE-Cooled Long-Life Safe Simple Small Portable Proliferation-Resistant Reactor) has been designed and continues to be improved. In the present paper a recent version of LSPR is presented. The total power of the present design is 150 MWt (53 MWe). During whole reactor life of 12 years the excess reactivity required for burnup is very low, and negative coolant dilatation coefficient is confirmed. This characteristic together with some other characteristics makes unprotected loss of flow (ULOF) accident inherently safe. It can survive even simultaneous rod run-out transient over power (UTOP), ULOF and unprotected loss of heat sink (ULOHS) accident without the help of an operator or active device. (authors)

  13. Structural, electrical and magnetic measurements on oxide layers grown on 316L exposed to liquid lead-bismuth eutectic

    NASA Astrophysics Data System (ADS)

    Hosemann, Peter; Hofer, Christian; Hlawacek, Gregor; Li, Ning; Maloy, Stuart A.; Teichert, Christian

    2012-02-01

    Fast reactors and spallation neutron sources may use lead-bismuth eutectic (LBE) as a coolant. Its physical, chemical, and irradiation properties make it a safe coolant compared to Na cooled designs. However, LBE is a corrosive medium for most steels and container materials. The present study was performed to evaluate the corrosion behavior of the austenitic steel 316L (in two different delivery states). Detailed atomic force microscopy, magnetic force microscopy, conductive atomic force microscopy, and scanning transmission electron microscopy analyses have been performed on the oxide layers to get a better understanding of the corrosion and oxidation mechanisms of austenitic and ferritic/martensitic stainless steel exposed to LBE. The oxide scale formed on the annealed 316L material consisted of multiple layers with different compositions, structures, and properties. The innermost oxide layer maintained the grain structure of what used to be the bulk steel material and shows two phases, while the outermost oxide layer possessed a columnar grain structure.

  14. LFR "Lead-Cooled Fast Reactor"

    SciTech Connect

    Cinotti, L; Fazio, C; Knebel, J; Monti, S; Abderrahim, H A; Smith, C; Suh, K

    2006-05-11

    development already being carried out in different institutes participating in this STREP. This is particularly true in Europe where a large R&D program associated with the development of Accelerator Driven Systems (ADS) is being actively pursued. The general objective of the ELSY project is to design an innovative lead-cooled fast reactor complemented by an analytical effort to assess the existing knowledge base in the field of lead-alloy coolants (i.e., lead-bismuth eutectic (LBE) and also lead/lithium) in order to extrapolate this knowledge base to pure lead. This analysis effort will be complemented with some limited R&D activities to acquire missing or confirmatory information about fundamental topics for ELSY that are not sufficiently covered in the ongoing European ADS program or elsewhere.

  15. Experiment and Numerical Simulation of Bubble Behavior in Argon Gas Injection into Lead-Bismuth Pool

    NASA Astrophysics Data System (ADS)

    Yamada, Yumi; Akashi, Toyou; Takahashi, Minoru

    In a lead-bismuth alloy (45%Pb-55%Bi) cooled direct contact boiling water fast reactor (PBWFR), steam can be produced by direct contact of feed water with primary Pb-Bi coolant in the upper core plenum, and Pb-Bi coolant can be circulated by buoyancy forces of steam bubbles. As a basic study to investigate the two-phase flow characteristics in the chimneys of PBWFR, a two-dimensional two-phase flow was simulated by injecting argon gas into Pb-Bi pool in a rectangular vessel (400mm in length, 1500mm in height), and bubble behavior were investigated experimentally. Bubble sizes, bubble rising velocities and void fractions were measured using void probes. The experimental conditions are the atmospheric pressure and the flow rate of injection Ar gas is 10, 20, and 30 NL/min. The average of measured bubble rising velocity was about 0.6 m/s. The average chord length was about 7mm. An analysis was performed by two-dimensional and two-fluid model. The experimental results were compared with the analytical results to evaluate the validity of the analytical model. Although large diameter bubbles were observed in the experiment, the drag force model of lower value performed better for simulation of the experimental result.

  16. Study of iron structure stability in high temperature molten lead-bismuth eutectic with oxygen injection using molecular dynamics simulation

    SciTech Connect

    Arkundato, Artoto; Su'ud, Zaki; Sudarko; Shafii, Mohammad Ali; Celino, Massimo

    2014-09-30

    Corrosion of structural materials in high temperature molten lead-bismuth eutectic is a major problem for design of PbBi cooled reactor. One technique to inhibit corrosion process is to inject oxygen into coolant. In this paper we study and focus on a way of inhibiting the corrosion of iron using molecular dynamics method. For the simulation results we concluded that effective corrosion inhibition of iron may be achieved by injection 0.0532 wt% to 0.1156 wt% oxygen into liquid lead-bismuth. At this oxygen concentration the structure of iron material will be maintained at about 70% in bcc crystal structure during interaction with liquid metal.

  17. Oxide layer stability in lead-bismuth at high temperature

    NASA Astrophysics Data System (ADS)

    Martín, F. J.; Soler, L.; Hernández, F.; Gómez-Briceño, D.

    2004-11-01

    Materials protection by 'in situ' oxidation has been studied in stagnant lead-bismuth, with different oxygen levels (H 2/H 2O ratios of 0.3 and 0.03), at temperatures from 535 °C to 600 °C and times from 100 to 3000 h. The materials tested were the martensitic steels F82Hmod, EM10 and T91 and the austenitic stainless steels, AISI 316L and AISI 304L. The results obtained point to the existence of an apparent threshold temperature above which corrosion occurs and the formation of a protective and stable oxide layer is not possible. This threshold temperature depends on material composition, oxygen concentration in the liquid lead-bismuth and time. The threshold temperature is higher for the austenitic steels, especially for the AISI 304L, and it increases with the oxygen concentration in the lead-bismuth. The oxide layer formed disappear with time and, after 3000 h all the materials, except AISI 304L, suffer corrosion, more severe for the martensitic steels and at the highest temperature tested.

  18. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    NASA Astrophysics Data System (ADS)

    Afifah, Maryam; Miura, Ryosuke; Su'ud, Zaki; Takaki, Naoyuki; Sekimoto, H.

    2015-09-01

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don't need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  19. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    SciTech Connect

    Afifah, Maryam Su’ud, Zaki; Miura, Ryosuke; Takaki, Naoyuki; Sekimoto, H.

    2015-09-30

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  20. Layer formation on metal surfaces in lead-bismuth at high temperatures in presence of zirconium

    NASA Astrophysics Data System (ADS)

    Loewen, Eric P.; Yount, Hannah J.; Volk, Kevin; Kumar, Arvind

    2003-09-01

    If the operating temperature lead-bismuth cooled fission reactor could be extended to 800 °C, they could produce hydrogen directly from water. A key issue for the deployment of this technology at these temperatures is the corrosion of the fuel cladding and structural materials by the lead-bismuth. Corrosion studies of several metals were performed to correlate the interaction layer formation rate as a function of time, temperature, and alloy compositions. The interaction layer is defined as the narrow band between the alloy substrate and the solidified lead-bismuth eutectic on the surface. Coupons of HT-9, 410, 316L, and F22 were tested at 550 and 650 °C for 1000 h inside a zirconium corrosion cell. The oxygen potential ranged from approximately 10 -22 to 10 -19 Pa. Analyses were performed on the coupons to determine the depth of the interaction layer and the composition, at each time step (100, 300, and 1000 h). The thickness of the interaction layer on F22 at 550 °C was 25.3 μm, the highest of all the alloys tested, whereas at 650 °C, the layer thickness was only 5.6 μm, the lowest of all the alloys tested. The growth of the interaction layer on F22 at 650 °C was suppressed, owing to the presence of Zr (at 1500 wppm) in the LBE. In the case of 316L, the interaction layers of 4.9 and 10.6 μm were formed at 550 and 650 °C, respectively.

  1. Thermodynamic assessment of solubility and activity of iron, chromium, and nickel in lead bismuth eutectic

    NASA Astrophysics Data System (ADS)

    Gossé, Stéphane

    2014-06-01

    Lead-Bismuth Eutectic (LBE) is a heavy liquid alloy used as a coolant for the Lead-Cooled Fast Reactors and spallation target for Accelerator Driven Systems. LBE is also considered in sodium fast reactor designs as coolant in secondary circuit to avoid any occurrence of the reaction between sodium and water in steam generators. Even if this coolant presents many advantages due to its thermophysical properties, corrosion towards structural materials remains one of the major issues of LBE. Because corrosion in LBE is partly driven by dissolution processes, the solubility and chemical activity of the main elements of the alloy are the key parameters to model the related corrosion processes. Using the Calphad method and the Thermo-Calc software, a thermodynamic database was developed to assess the interaction between Cr-Ni-Fe alloys and LBE. The current thermodynamic data on the Cr-Fe-Ni + Bi-Pb quinary system was reviewed and the Bi-Cr and Cr-Pb binary phase diagrams were assessed. Fe, Cr and Ni solubilities (in at. fraction, T in K) at LBE composition were calculated: Fe solubility at LBE composition: log10 (SFe)=0.5719-4398.6T (399-1173 K) Cr solubility at LBE composition: log10 (SCr)=-0.2757-3056.1T (399-1173 K) Ni solubility at LBE composition: log10 (SNi)=2.8717-2932.9T (528-742 K) log10 (SNi)=0.2871-1006.3T (742-1173 K) Then, the thermodynamic assessment performed in this study was used to predict more accurately the Fe, Cr and Ni activities and solubilities in the case of four austenitic model alloys also studied in the framework of corrosion tests [1]. The calculated activities and solubilities provide thermodynamic data to better understand dissolution or precipitation phenomena observed during LBE corrosion processes.

  2. Ni-rich precipitates in a lead bismuth eutectic loop

    NASA Astrophysics Data System (ADS)

    Kikuchi, K.; Saito, S.; Hamaguchi, D.; Tezuka, M.

    2010-03-01

    Solidified LBE was sampled from the specimens, electro-magnetic pump, filter, drain valve and oxygen sensor at the JAEA Lead Bismuth Loop-1 (JLBL-1) where the structural material was made of SS316. The concentration of Ni, Fe and Cr in LBE were analyzed by the Inductive Coupled Plasma atomic emission spectrometer. It was concluded that the solution of Ni into LBE was not saturated although the concentration of Fe and Cr almost achieved to the values in the literature. A needle-type structure appeared on the surface of solidified LBE inside the tube specimens. It was found to be Ni-rich precipitates by X-ray analyses (Field Emission Scanning Electron Microscope, FE-SEM). LBE samples collected from a circulating loop after discharging did not show the amount of impurities equivalent to the LBE bulk property.

  3. Oxygen concentration measurement in liquid lead-bismuth eutectic

    SciTech Connect

    Darling, T. W.; Li, N.

    2001-01-01

    Liquid lead-bismuth (Pb-Bi) eutectic (LBE) may see extensive use as a coolant fluid, and perhaps also as a spallation target, in next generation nuclear energy systems. While it is not as reactive as alkali metal liquids, it does present a long term corrosion problem with some materials, notably stainless steels. Mitigation of the corrosion problem may be achieved by producing and maintaining a protective oxide on exposed surfaces, through control of the concentration of dissolved oxygen in the LBE. We have developed an oxygen sensor based on available zirconia-based solid electrolytes used in the automotive industry, which represents a relatively inexpensive source of reproducible and reliable components. We will present the design considerations and characteristics of our sensor unit, and describe its use in the LBE test loop at Los Alamos for measurement and control of dissolved oxygen concentration.

  4. Evaporation of mercury impurity from liquid lead-bismuth eutectic

    NASA Astrophysics Data System (ADS)

    Aerts, A.; Danaci, S.; Gonzalez Prieto, B.; Van den Bosch, J.; Neuhausen, J.

    2014-05-01

    The equilibrium evaporation of mercury from dilute solutions in liquid lead-bismuth eutectic (LBE) was studied in argon atmosphere. Mercury present as impurity in LBE was evaporated and detected by atomic fluorescence spectroscopy. A method which could accurately simulate the experimental data was developed. Coefficients of the Henry constant temperature correlation for mercury dissolved in LBE were determined. Experiments with samples from several different batches of LBE revealed that mercury at mole fractions between 10-6 and 10-12 and temperatures between 150 and 350 °C evaporated from liquid LBE close to ideal behavior. Evaporation of mercury from solid LBE on the other hand was unexpectedly high. These results are important for safety evaluations of LBE based spallation targets and accelerator driven systems.

  5. Modeling astatine production in liquid lead-bismuth spallation targets

    NASA Astrophysics Data System (ADS)

    David, J. C.; Boudard, A.; Cugnon, J.; Ghali, S.; Leray, S.; Mancusi, D.; Zanini, L.

    2013-03-01

    Astatine isotopes can be produced in liquid lead-bismuth eutectic targets through proton-induced double charge exchange reactions on bismuth or in secondary helium-induced interactions. Models implemented into the most common high-energy transport codes generally have difficulties to correctly estimate their production yields as was shown recently by the ISOLDE Collaboration, which measured release rates from a lead-bismuth target irradiated by 1.4 and 1 GeV protons. In this paper, we first study the capability of the new version of the Liège intranuclear cascade model, INCL4.6, coupled to the deexcitation code ABLA07 to predict the different elementary reactions involved in the production of such isotopes through a detailed comparison of the model with the available experimental data from the literature. Although a few remaining deficiencies are identified, very satisfactory results are found, thanks in particular to improvements brought recently on the treatment of low-energy helium-induced reactions. The implementation of the models into MCNPX allows identifying the respective contributions of the different possible reaction channels in the ISOLDE case. Finally, the full simulation of the ISOLDE experiment is performed, taking into account the likely rather long diffusion time from the target, and compared with the measured diffusion rates for the different astatine isotopes, at the two studied energies, 1.4 and 1 GeV. The shape of the isotopic distribution is perfectly reproduced as well as the absolute release rates, assuming in the calculation a diffusion time between 5 and 10hours. This work finally shows that our model, thanks to the attention paid to the emission of high-energy clusters and to low-energy cluster induced reactions, can be safely used within MCNPX to predict isotopes with a charge larger than that of the target by two units in spallation targets, and, probably, more generally to isotopes created in secondary reactions induced by composite

  6. Polonium problem in lead-bismuth flow target

    SciTech Connect

    Pankratov, D.V.; Yefimov, E.I.; Bugreev, M.I.

    1996-06-01

    Alpha-active polonium nuclides Po198 - Po210 are formed in a lead-bismuth target as results of reactions Bi{sup 209}(n,{gamma})Bi{sup 210} {yields} Po{sup 210}, Bi{sup 209}(p,xn)Po{sup 210} {yields} Po{sup 210 {minus} x} (x = 1-12), Pb{sup 208}({alpha},xn) {yields} Po{sup 210 {minus} x + 2} (x = 2-14). The most important nuclides are Po-210 (T{sub {1/2}}=138.4 day), Po-209 (T{sub {1/2}}=102 years) and Po-208 (T{sub {1/2}}=2.9 years). Polonium activity of the circuit for SINQ - conditions is about 15,000 Ci after 1-year operation. Polonium radiation hazard is connected with its output from the coolant and formation of aerosol and surface alpha-activity after the circuit break-down for repair works or in accidents. One of the important issues of polonium removal system creation is containing and storing polonium removed. Its storage in solidified alkaline is not expedient because of secondary neutron formation as a result of ({alpha},n) - reaction on oxygen and sodium nucleus. The estimations carried out demonstrated that by polonium concentration {approx} 100 Ci/l neutron current on the container surface can reach {approx} 10{sup 4}n/(cm{sup 2}s). Concentration and storage of polonium in solidified lead-bisumth seems the most convenient. The calculations demonstrated that in a 100 l container 50,000 Ci of polonium can be stored (as much as 3 times more than 1-year polonium product in SINQ-conditions) under temperature in the container less than melting point of lead bismuth (the wall temperature is about 100{degrees}C).

  7. The Chemical Kinetics of Alkaline Extraction of Tellurium from Lead-Bismuth Eutectic

    SciTech Connect

    Laurence E. Auman; Eric P. Loewen; Thomas F. Gesell; Shuji Ohno

    2005-07-01

    Polonium-210 is an important radioactive product of neutron activation of molten lead-bismuth eutectic, a promising candidate coolant for advanced fast nuclear reactors. The radiological hazard potential associated with polonium can be significantly reduced by continuous online removal of polonium from the coolant. The removal method under investigation in this research is alkaline extraction. Chemical kinetic measurements were made to determine first and second order rate constants, activation energy, and heat of reaction at various temperatures using tellurium as a surrogate. First and second order alkaline extraction rate constants were measured to be: k1 = 10.05 e –52,274/RT and k2 = 167 e –97,224/RT. Alkaline extraction is dependent on temperature and was found to follow the Arrhenius rate law. The activation energy (Ea) ranged between 52,274 – 97,224 J mol-1. With a strong foundation of surrogate work completed, this work should be validated using polonium-210.

  8. Experimental investigation of forced-convection heat-transfer characteristics of lead-bismuth eutectic

    NASA Technical Reports Server (NTRS)

    Lubarsky, Bernard

    1951-01-01

    The forced-convection heat-transfer characteristics of lead-bismuth eutectic were experimentally investigated. Experimental values of Nusselt number for lead-bismuth fell considerably below predicted values. The addition of a wetting agent did not change the heat transfer characteristics.

  9. Corrosion by liquid lead and lead-bismuth: experimental results review and analysis

    SciTech Connect

    Zhang, Jinsuo

    2008-01-01

    Liquid metal technologies for liquid lead and lead-bismuth alloy are under wide investigation and development for advanced nuclear energy systems and waste transmutation systems. Material corrosion is one of the main issues studied a lot recently in the development of the liquid metal technology. This study reviews corrosion by liquid lead and lead bismuth, including the corrosion mechanisms, corrosion inhibitor and the formation of the protective oxide layer. The available experimental data are analyzed by using a corrosion model in which the oxidation and scale removal are coupled. Based on the model, long-term behaviors of steels in liquid lead and lead-bismuth are predictable. This report provides information for the selection of structural materials for typical nuclear reactor coolant systems when selecting liquid lead or lead bismuth as heat transfer media.

  10. Experimental and numerical study on lead-bismuth heat transfer in a fuel rod simulator

    NASA Astrophysics Data System (ADS)

    Ma, Weimin; Karbojian, Aram; Hollands, Thorsten; Koch, Marco K.

    2011-08-01

    As a task of the EU project IP EUROTRANS towards development of an Accelerator Driven System (ADS) dedicated to the transmutation of long-lived fission products, experiments and simulations were performed on the TALL test facility at KTH to investigate thermal hydraulics along a single fuel rod simulator cooled by lead-bismuth eutectic (LBE). The fuel rod simulator is concentrically inserted in a tube, so that an annular channel is formed for LBE flow. This paper presents the measured temperature profiles in the annular channel, and the comparisons with the simulation results of the CFX code. The primary objective is to help understanding the LBE heat transfer characteristics and qualifying the turbulence and heat transfer modeling for LBE application. The quantitative comparison between the calculated and measured temperatures of the LBE indicates that the simulation underestimates the experiment at most radial and axial positions. Finally the uncertainties in measurement and the deficiency in turbulence models resulting in such a disagreement were discussed, which will be directive and beneficial to future work in the field.

  11. Boosted Fast Flux Loop Alternative Cooling Assessment

    SciTech Connect

    Glen R. Longhurst; Donna Post Guillen; James R. Parry; Douglas L. Porter; Bruce W. Wallace

    2007-08-01

    The Gas Test Loop (GTL) Project was instituted to develop the means for conducting fast neutron irradiation tests in a domestic radiation facility. It made use of booster fuel to achieve the high neutron flux, a hafnium thermal neutron absorber to attain the high fast-to-thermal flux ratio, a mixed gas temperature control system for maintaining experiment temperatures, and a compressed gas cooling system to remove heat from the experiment capsules and the hafnium thermal neutron absorber. This GTL system was determined to provide a fast (E > 0.1 MeV) flux greater than 1.0E+15 n/cm2-s with a fast-to-thermal flux ratio in the vicinity of 40. However, the estimated system acquisition cost from earlier studies was deemed to be high. That cost was strongly influenced by the compressed gas cooling system for experiment heat removal. Designers were challenged to find a less expensive way to achieve the required cooling. This report documents the results of the investigation leading to an alternatively cooled configuration, referred to now as the Boosted Fast Flux Loop (BFFL). This configuration relies on a composite material comprised of hafnium aluminide (Al3Hf) in an aluminum matrix to transfer heat from the experiment to pressurized water cooling channels while at the same time providing absorption of thermal neutrons. Investigations into the performance this configuration might achieve showed that it should perform at least as well as its gas-cooled predecessor. Physics calculations indicated that the fast neutron flux averaged over the central 40 cm (16 inches) relative to ATR core mid-plane in irradiation spaces would be about 1.04E+15 n/cm2-s. The fast-to-thermal flux ratio would be in excess of 40. Further, the particular configuration of cooling channels was relatively unimportant compared with the total amount of water in the apparatus in determining performance. Thermal analyses conducted on a candidate configuration showed the design of the water coolant and

  12. Numerical study: Iron corrosion-resistance in lead-bismuth eutectic coolant by molecular dynamics method

    SciTech Connect

    Arkundato, Artoto; Su'ud, Zaki; Abdullah, Mikrajuddin; Widayani,; Celino, Massimo

    2012-06-06

    In this present work, we report numerical results of iron (cladding) corrosion study in interaction with lead-bismuth eutectic coolant of advanced nuclear reactors. The goal of this work is to study how the oxygen can be used to reduce the corrosion rate of cladding. The molecular dynamics method was applied to simulate corrosion process. By evaluating the diffusion coefficients, RDF functions, MSD curves of the iron and also observed the crystal structure of iron before and after oxygen injection to the coolant then we concluded that a significant and effective reduction can be achieved by issuing about 2% number of oxygen atoms to lead-bismuth eutectic coolant.

  13. Safety Characteristics of LBE Cooled Long-Life Small Reactor, 'LSPR'

    SciTech Connect

    Hiroshi Sekimoto; Shinichi Makino

    2002-07-01

    Lead bismuth eutectic (LBE) shows a good performance on neutron economy, and LBE cooled fast reactor can be designed as an excellent long-life small reactor. LBE is good not only for neutron economy but for chemical inertness and high boiling point, which may realize a much safer reactor than conventional sodium-cooled reactor. We have designed such a long-life small reactor and name it LSPR. This paper presents safety characteristics of LSPR. (authors)

  14. Characterization of oxide layers grown on D9 austenitic stainless steel in lead bismuth eutectic

    NASA Astrophysics Data System (ADS)

    Hosemann, P.; Hawley, M.; Koury, D.; Swadener, J. G.; Welch, J.; Johnson, A. L.; Mori, G.; Li, N.

    2008-04-01

    Lead bismuth eutectic (LBE) is a possible coolant for fast reactors and targets in spallation neutron sources. Its low melting point, high evaporation point, good thermal conductivity, low reactivity, and good neutron yield make it a safe and high performance coolant in radiation environments. The disadvantage is that it is a corrosive medium for most steels and container materials. This study was performed to evaluate the corrosion behavior of the austenitic stainless steel D9 in oxygen controlled LBE. In order to predict the corrosion behavior of steel in this environment detailed analyses have to be performed on the oxide layers formed on these materials and various other relevant materials upon exposure to LBE. In this study the corrosion/oxidation of D9 stainless steel in LBE was investigated in great detail. The oxide layers formed were characterized using atomic force microscopy, magnetic force microscopy, nanoindentation, and scanning electron microscopy with wavelength-dispersive spectroscopy (WDS) to understand the corrosion and oxidation mechanisms of D9 stainless steel in contact with the LBE. What was previously believed to be a simple double oxide layer was identified here to consist of at least 4 different oxide layers. It was found that the inner most oxide layer takes over the grain structure of what used to be the bulk steel material while the outer oxide layer consists of freshly grown oxides with a columnar structure. These results lead to a descriptive model of how these oxide layers grow on this steel under the harsh environments encountered in these applications.

  15. Problems of development the pilot lead-bismuth target circuit TC-1 for ADS

    SciTech Connect

    Ignatiev, Sviatoslav; Leonchuk, Mikhail; Orlov, Yury; Pankratov, Dmitry; Suvorov, Gennady; Zabudko, Alexey

    2007-07-01

    Available in abstract form only. Full text of publication follows: Main problems on development of pilot molten lead-bismuth target circuit of 1 MW proton beam power (TC-1) as an important part of target-blanket accelerator driven system (ADS) for nuclear waste incineration are analyzed. (authors)

  16. Oxygen-iron interaction in liquid lead-bismuth eutectic alloy.

    PubMed

    Aerts, A; Gavrilov, S; Manfredi, G; Marino, A; Rosseel, K; Lim, J

    2016-07-20

    Iron released by steel corrosion was found to be a key impurity in reactions with dissolved oxygen in liquid lead-bismuth eutectic alloys. The iron-oxygen-magnetite equilibrium was characterized, allowing the quantification of phenomena that are important for long-term operation of lead-alloy based installations such as corrosion rate control and management of precipitates. PMID:27383127

  17. A resting bottom sodium cooled fast reactor

    SciTech Connect

    Costes, D.

    2012-07-01

    This follows ICAPP 2011 paper 11059 'Fast Reactor with a Cold Bottom Vessel', on sodium cooled reactor vessels in thermal gradient, resting on soil. Sodium is frozen on vessel bottom plate, temperature increasing to the top. The vault cover rests on the safety vessel, the core diagrid welded to a toric collector forms a slab, supported by skirts resting on the bottom plate. Intermediate exchangers and pumps, fixed on the cover, plunge on the collector. At the vessel top, a skirt hanging from the cover plunges into sodium, leaving a thin circular slit partially filled by sodium covered by argon, providing leak-tightness and allowing vessel dilatation, as well as a radial relative holding due to sodium inertia. No 'air conditioning' at 400 deg. C is needed as for hanging vessels, and this allows a large economy. The sodium volume below the slab contains isolating refractory elements, stopping a hypothetical corium flow. The small gas volume around the vessel limits any LOCA. The liner cooling system of the concrete safety vessel may contribute to reactor cooling. The cold resting bottom vessel, proposed by the author for many years, could avoid the complete visual inspection required for hanging vessels. However, a double vessel, containing support skirts, would allow introduction of inspecting devices. Stress limiting thermal gradient is obtained by filling secondary sodium in the intermediate space. (authors)

  18. The Industrial Sodium Cooled Fast Reactor

    SciTech Connect

    Samuel E. Bays; Haihua Zhao; Hongbin Zhang

    2009-04-01

    This paper investigates the use of enrichment and moderator zoning methods for optimizing the r-z power distribution within sodium cooled fast reactors. These methods allow overall greater fuel utilization in the core resulting in more fuel being irradiated near the maximum allowed thermal power. The peak-to-average power density was held to 1.18. This core design, in conjunction with a multiple-reheat Brayton power conversion system, has merit for producing an industrial level of electrical output (2400MWth, 1000MWe) from a relatively compact core size. The total core radius, including reflectors and shields, was held to 1.78m. Preliminary safety analysis suggests that positive reactivity insertion resulting from a leak between the sodium primary loop and helium power conversion system can be mitigated using simple gas-liquid centripetal separation strategies in the plant’s primary loop.

  19. Idaho National Laboratory Lead or Lead-Bismuth Eutectic (LBE) Test Facility - R&D Requirements, Design Criteria, Design Concept, and Concept Guidance

    SciTech Connect

    Eric P. Loewen; Paul Demkowicz

    2005-05-01

    The Idaho National Laboratory Lead-Bismuth Eutectic Test Facility will advance the state of nuclear technology relative to heavy-metal coolants (primarily Pb and Pb-Bi), thereby allowing the U.S. to maintain the pre-eminent position in overseas markets and a future domestic market. The end results will be a better qualitative understanding and quantitative measure of the thermal physics and chemistry conditions in the molten metal systems for varied flow conditions (single and multiphase), flow regime transitions, heat input methods, pumping requirements for varied conditions and geometries, and corrosion performance. Furthering INL knowledge in these areas is crucial to sustaining a competitive global position. This fundamental heavy-metal research supports the National Energy Policy Development Group’s stated need for energy systems to support electrical generation.1 The project will also assist the Department of Energy in achieving goals outlined in the Nuclear Energy Research Advisory Committee Long Term Nuclear Technology Research and Development Plan,2 the Generation IV Roadmap for Lead Fast Reactor development, and Advanced Fuel Cycle Initiative research and development. This multi-unit Lead-Bismuth Eutectic Test Facility with its flexible and reconfigurable apparatus will maintain and extend the U.S. nuclear knowledge base, while educating young scientists and engineers. The uniqueness of the Lead-Bismuth Eutectic Test Facility is its integrated Pool Unit and Storage Unit. This combination will support large-scale investigation of structural and fuel cladding material compatibility issues with heavy-metal coolants, oxygen chemistry control, and thermal hydraulic physics properties. Its ability to reconfigure flow conditions and piping configurations to more accurately approximate prototypical reactor designs will provide a key resource for Lead Fast Reactor research and development. The other principal elements of the Lead-Bismuth Eutectic Test Facility

  20. A Comparison of Long-Lived, Proliferation Resistant Fast Reactors

    SciTech Connect

    Weaver, Kevan Dean; Herring, James Stephen; Mac Donald, Philip Elsworth

    2001-09-01

    Various methods have been proposed to transmute and thus consume the current inventory of trans-uranic waste that exists in spent light-water-reactor fuel. These methods include both critical and sub-critical systems. The neutronics of metallic and nitride fuels loaded with 20-30wt% light-water-reactor plutonium plus minor actinides for use in a lead-bismuth and sodium cooled fast reactor are discussed, with an emphasis on the fuel cycle life and isotopic content. Calculations show that core life can extend beyond 20 years, and the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from 0.5 to 0.9 g/MWd.

  1. Optimized sympathetic cooling of atomic mixtures via fast adiabatic strategies

    SciTech Connect

    Choi, Stephen; Sundaram, Bala; Onofrio, Roberto

    2011-11-15

    We discuss fast frictionless cooling techniques in the framework of sympathetic cooling of cold atomic mixtures. It is argued that optimal cooling of an atomic species--in which the deepest quantum degeneracy regime is achieved--may be obtained by means of sympathetic cooling with another species whose trapping frequency is dynamically changed to maintain constancy of the Lewis-Riesenfeld adiabatic invariant. Advantages and limitations of this cooling strategy are discussed, with particular regard to the possibility of cooling Fermi gases to a deeper degenerate regime.

  2. Polonium release from an ATW burner system with liquid lead-bismuth coolant

    SciTech Connect

    Li, N.; Yefimov, E.; Pankratov, D.

    1998-04-01

    The authors analyzed polonium release hazards in a conceptual pool-type ATW burner with liquid lead-bismuth eutectic (LBE) coolant. Simplified quantitative models are used based on experiments and real NPP experience. They found little Po contamination outside the burner under normal operating conditions with nominal leakage from the gas system. In sudden gas leak and/or coolant spill accidents, the P contamination level can reach above the regulation limit but short exposure would not lead to severe health consequences. They are evaluating and developing mitigation methods.

  3. Real Time Corrosion Monitoring in Lead and Lead-Bismuth Systems

    SciTech Connect

    James F. Stubbins; Alan Bolind; Ziang Chen

    2010-02-25

    The objective of this research program is to develop a real-time, in situ corrosion monitoring technique for flowing liquid Pb and eutectic PbBi (LBE) systems in a temperature range of 400 to 650 C. These conditions are relevant to future liquid metal cooled fast reactor operating parameters. THis program was aligned with the Gen IV Reactor initiative to develp technologies to support the design and opertion of a Pb or LBE-cooled fast reactor. The ability to monitor corrosion for protection of structural components is a high priority issue for the safe and prolonged operation of advanced liquid metal fast reactor systems. In those systems, protective oxide layers are intentionally formed and maintained to limit corrosion rates during operation. This program developed a real time, in situ corrosion monitoring tecnique using impedance spectroscopy (IS) technology.

  4. Potential containment materials for liquid-lead and lead-bismuth eutectic spallation neutron source

    SciTech Connect

    Park, J.J.; Butt, D.P.; Beard, C.A.

    1997-11-01

    Lead (Pb) and lead-bismuth eutectic (44Pb-56Bi) have been the two primary candidate liquid-metal target materials for the production of spallation neutrons. Selection of a container material for the liquid-metal target will greatly affect the lifetime and safety of the target subsystem. For the lead target, niobium-1 (wt%) zirconium (Nb-1Zr) is a candidate containment material for liquid lead, but its poor oxidation resistance has been a major concern. The oxidation rate of Nb-1Zr was studied based on the calculations of thickness loss due to oxidation. According to these calculations, it appeared that uncoated Nb-1Zr may be used for a one-year operation at 900 C at P{sub O{sub 2}} = 1 {times} 10{sup {minus}6} torr, but the same material may not be used in argon with 5-ppm oxygen. Coating technologies to reduce the oxidation of Nb-1Zr are reviewed, as are other candidate refractory metals such as molybdenum, tantalum, and tungsten. For the Pb-Bi target, three candidate containment materials are suggested based on a literature survey of the materials compatibility and proton irradiation tests: Croloy 2-1/4, modified 9Cr-1Mo, and 12Cr-1Mo (HT-9) steel. These materials seem to be used only if the lead-bismuth is thoroughly deoxidized and treated with zirconium and magnesium.

  5. Effects of temperature and strain rate on the tensile behaviors of SIMP steel in static lead bismuth eutectic

    NASA Astrophysics Data System (ADS)

    Liu, Jian; Yan, Wei; Sha, Wei; Wang, Wei; Shan, Yiyin; Yang, Ke

    2016-05-01

    In order to assess the susceptibility of candidate structural materials to liquid metal embrittlement, this work investigated the tensile behaviors of ferritic-martensitic steel in static lead bismuth eutectic (LBE). The tensile tests were carried out in static lead bismuth eutectic under different temperatures and strain rates. Pronounced liquid metal embrittlement phenomenon is observed between 200 °C and 450 °C. Total elongation is reduced greatly due to the liquid metal embrittlement in LBE environment. The range of ductility trough is larger under slow strain rate tensile (SSRT) test.

  6. Natural Circulation of Lead-Bismuth in a One-Dimensional Loop: Experiments and Code Predictions

    SciTech Connect

    Agostini, P.; Bertacci, G.; Gherardi, G.; Bianchi, F.; Meloni, P.; Nicolini, D.; Ambrosini, W.; Forgione, F.; Fruttuoso, G.; Oriolo, F.

    2002-07-01

    The paper summarizes the results obtained by an experimental and computational study jointly performed by ENEA and University of Pisa. The study is aimed at assessing the capabilities of an available thermal-hydraulic system code in simulating natural circulation in a loop in which the working fluid is the eutectic lead-bismuth alloy as in the Italian proposal for Accelerator Driven System (ADS) reactor concepts. Experiments were performed in the CHEOPE facility installed at the ENEA Brasimone Research Centre and pre- and post-test calculations were run using a version of the RELAP5/Mod.3.2, purposely modified to account for Pb-Bi liquid alloy properties and behavior. The main results obtained by the experimental tests and by the code analyses are presented in the paper providing material to discuss the present predictive capabilities of transient and steady-state behavior in liquid Pb-Bi systems. (authors)

  7. Parametric study of a corrosion model applied to lead-bismuth flow systems

    NASA Astrophysics Data System (ADS)

    Zhang, Jinsuo; Li, Ning

    2003-09-01

    The corrosion of steels exposed to flowing liquid metals is influenced by local and axial conditions of the flow systems. Despite of this, most existing corrosion models only consider the mean values based on local conditions. The present study refines a model for flowing liquid metal under non-isothermal conditions. The model is based on solving the mass transport equation in the boundary layer. Two kinds of flows are investigated: through an open pipe system and through a closed loop system. The model is applied to a lead-bismuth eutectic (LBE) test loop. A parametric study illustrates the effects of the axial temperature profile on corrosion. The study provides important insight to the design, operation and testing of such loop systems.

  8. Interim status report on lead-cooled fast reactor (LFR) research and development.

    SciTech Connect

    Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.; Smith, C. F.; de Caro, M.; Halsey, W. G.; Li, N.; Hosemann, P.; Zhang, J.; Bolind, A.; LLNL; LANL; Univ. of Illinois

    2008-03-31

    15} (n/cm{sup 2}-s) and the initially 563 MWt PHENIX reactor attained 2.0 x 10{sup 15} (n/cm{sup 2}-s) before one of three intermediate cooling loops was shut down due to concerns about potential steam generator tube failures. The calculations do not assume a test assembly location for advanced fuels and materials irradiation in place of a fuel assembly (e.g., at the center of the core); the calculations have not examined whether it would be feasible to replace the central assembly by a test assembly location. However, having only fifteen driver assemblies implies a significant effect due to perturbations introduced by the test assembly. The peak neutron fast flux is low compared with the fast fluxes previously achieved in FFTF and PHENIX. Furthermore, the peak neutron fluence is only about half of the limiting value (4 x 10{sup 23} n/cm{sup 2}) typically used for ferritic steels. The results thus suggest that a larger power level (e.g., 400 MWt) and a larger core would be better for a TPP based upon the ELSY fuel assembly design and which can also perform irradiation testing of advanced fuels and materials. In particular, a core having a higher power level and larger dimensions would achieve a suitable average discharge burnup, peak fast flux, peak fluence, and would support the inclusion of one or more test assembly locations. Participation in the Generation IV International Forum Provisional System Steering Committee for the LFR is being maintained throughout FY 2008. Results from the analysis of samples previously exposed to flowing lead-bismuth eutectic (LBE) in the DELTA loop are summarized and a model for the oxidation/corrosion kinetics of steels in heavy liquid metal coolants was applied to systematically compare the calculated long-term (i.e., following several years of growth) oxide layer thicknesses of several steels.

  9. Experimental Investigation of Evaporation Behavior of Polonium and Rare-Earth Elements in Lead-Bismuth Eutectic Pool

    SciTech Connect

    Shuji Ohno; Shinya Miyahara; Yuji Kurata; Ryoei Katsura; Shigeru Yoshida

    2006-07-01

    Equilibrium evaporation behavior was experimentally investigated for polonium ({sup 210}Po) in liquid lead-bismuth eutectic (LBE) and for rare-earth elements gadolinium (Gd) and europium (Eu) in LBE to understand and clarify the transfer behavior of toxic impurities from LBE coolant to a gas phase. The experiments utilized the 'transpiration method' in which saturated vapor in an isothermal evaporation pot was transported by inert carrier gas and collected outside of the pot. While the previous paper ICONE12-49111 has already reported the evaporation behavior of LBE and of tellurium in LBE, this paper summarizes the outlines and the results of experiments for important impurity materials {sup 210}Po and rare-earth elements which are accumulated in liquid LBE as activation products and spallation products. In the experiments for rare-earth elements, non-radioactive isotope was used. The LBE pool is about 330-670 g in weight and has a surface area of 4 cm x 14 cm. {sup 210}Po experiments were carried out with a smaller test apparatus and radioactive {sup 210}Po produced through neutron irradiation of LBE in the Japan Materials Testing Reactor (JMTR). We obtained fundamental and instructive evaporation data such as vapor concentration, partial vapor pressure of {sup 210}Po in the gas phase, and gas-liquid equilibrium partition coefficients of the impurities in LBE under the temperature condition between 450 and 750 deg. C. The {sup 210}Po test revealed that Po had characteristics to be retained in LBE but was still more volatile than LBE solvent. A part of Eu tests implied high volatility of rare-earth elements comparable to that of Po. This tendency is possibly related to the local enrichment of the solute near the pool surface and needs to be investigated more. These results are useful and indispensable for the evaluation of radioactive materials transfer to the gas phase in LBE-cooled nuclear systems. (authors)

  10. Experiments Performed in Substantiation of the Conditioning of BN-350 Spent Cesium Traps Using Lead or Lead-Bismuth Alloy Filling Technology

    SciTech Connect

    O. Romanenko; I. Tazhibaeva; I. Yakovlev; A. Ivanov; D. Wells; A. Herrick; J. Michelbacher; S. Shiganakov

    2009-05-01

    The technology of cleaning cesium radionuclides from sodium coolant at the BN-350 fast reactor was realized in the form of cesium traps of two types: stationary devices connected to the circuit that was to be cleaned and in-core devices installed into the core of reactor when it was not under operation. Carbon-graphite materials were used as sorbents to collect and concentrate radioactive cesium, accumulated in the BN-350 reactor circuits over the decades of their operation, in relatively small volume traps which provided effective radiation-safe conditions for personnel working in proximity to the coolant and equipment of the primary circuit during BN-350 decommissioning. Spent cesium traps, as products unfit for further use, represent solid radioactive wastes. The presence of chemically active sodium, potassium and cesium that are able to react violently with water results in series of problems related to their disposal in the Republic of Kazakhstan. Considering the technology of filling spent cesium traps with lead/lead-bismuth alloy as a priority one for their conditioning, evaluations for safety substantiation were implemented. A set of experiments was implemented aimed at verification of calculations performed in substantiation of the proposed technology: filling a full scale cesium trap mock-up with sodium followed by its draining to determine the optimal regimes of draining; filling bench scale cesium trap mock-ups with sodium and cesium followed by sodium draining and filling with lead or lead-bismuth alloy at different temperatures and filling rates to chose the optimal regimes for filling spent cesium traps; implementation of leachability tests to determine the rate of cesium release from the filling materials into water. This paper provides a description of the experimental program carried out and the main results obtained.

  11. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    SciTech Connect

    Hongbin Zhang; Haihua Zhao

    2009-05-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phénix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design.

  12. Compatibility of martensitic/austenitic steel welds with liquid lead bismuth eutectic environment

    NASA Astrophysics Data System (ADS)

    Van den Bosch, J.; Almazouzi, A.

    2009-04-01

    The high-chromium ferritic/martensitic steel T91 and the austenitic stainless steel 316L are to be used in contact with liquid lead-bismuth eutectic (LBE), under high irradiation doses. Both tungsten inert gas (TIG) and electron beam (EB) T91/316L welds have been examined by means of metallography, scanning electron microscopy (SEM-EDX), Vickers hardness measurements and tensile testing both in inert gas and in LBE. Although the T91/316L TIG weld has very good mechanical properties when tested in air, its properties decline sharply when tested in LBE. This degradation in mechanical properties is attributed to the liquid metal embrittlement of the 309 buttering used in TIG welding of T91/316L welds. In contrast to mixed T91/316L TIG welding, the mixed T91/316L EB weld was performed without buttering. The mechanical behaviour of the T91/316L EB weld was very good in air after post weld heat treatment but deteriorated when tested in LBE.

  13. Distribution and surface enrichment of radionuclides in lead-bismuth eutectic from spallation targets

    NASA Astrophysics Data System (ADS)

    Hammer-Rotzler, Bernadette; Neuhausen, Jörg; Boutellier, Viktor; Wohlmuther, Michael; Zanini, L.; David, J.-C.; Türler, Andreas; Schumann, Dorothea

    2016-07-01

    With the development of new high-power neutron spallation sources --both for scientific application and as neutron production tool for accelerator-driven systems-- the demand for experimentally obtained nuclear data on the residue nuclei production in the target is constantly increasing. In the present work, we examined two lead-bismuth-eutectic targets, irradiated with high-energy protons, concerning their radionuclide content and the spatial distribution of selected isotopes. The first one was the so-called ISOLDE target, being irradiated with 1-1.4GeV protons at CERN-ISOLDE, the second one was the MEGAPIE target, irradiated at PSI with 590MeV protons. In particular, we investigated the phenomenon of radionuclide enrichment on free surfaces in both targets. It turned out that considerable accumulation can be found especially in the case of lanthanides. The depletion process is enhanced at increased temperatures. The results are compared with theoretical predictions; some possible consequences of the findings are illustrated.

  14. Experimental and Analytical Study of Lead-Bismuth-Water Direct Contact Boiling Two-Phase Flow

    NASA Astrophysics Data System (ADS)

    Novitrian; Dostal, Vaclav; Takahashi, Minoru

    The characteristics of lead-bismuth(Pb-Bi)-water boiling two-phase flow were investigated experimentally and analytically using a Pb-Bi-water direct contact boiling two-phase flow loop. Pb-Bi flow rates and void fraction were measured in a vertical circular tube at conditions of system pressure 7MPa, liquid metal temperature 460°C and injected water temperature 220°C. The drift-flux model with the assumption that bubble sizes were dependent on the fluid surface tension and the density ratio of Pb-Bi to steam-water mixture was chosen and modified by the best fit to the measured void fraction. Pb-Bi flow rates were analytically estimated using balance condition between buoyancy force and pressure losses, where the buoyancy force was calculated from void fraction estimated using the modified drift-flux model. The deviation of the analytical results of the flow rates from the experimental ones was less than 10%.

  15. Interaction study between MOX fuel and eutectic lead-bismuth coolant

    NASA Astrophysics Data System (ADS)

    Vigier, Jean-François; Popa, Karin; Tyrpekl, Vaclav; Gardeur, Sébastien; Freis, Daniel; Somers, Joseph

    2015-12-01

    In the frame of the MYRRHA reactor project, the interaction between fuel pellets and the reactor coolant is essential for safety evaluations, e.g. in case of a pin breach. Therefore, interaction tests between uranium-plutonium mixed oxide (MOX) pellets and molten lead bismuth eutectic (LBE) have been performed and three parameters were studied, namely the interaction temperature (500 °C and 800 °C), the oxygen content in LBE and the stoichiometry of the MOX (U0.7Pu0.3O2-x and U0.7Pu0.3O2.00). After 50 h of interaction in closed containers, the pellet integrity was preserved in all cases. Whatever the conditions, neither interaction compounds (crystalline or amorphous) nor lead and bismuth diffusion into the surface regions of the MOX pellets has been detected. In most of the conditions, actinide releases into LBE were very limited (in the range of 0.01-0.15 mg), with a homogeneous release of the different actinides present in the MOX. Detected values were significantly higher in the 800 °C and low LBE oxygen content tests for both U0.7Pu0.3O2-x and U0.7Pu0.3O2.00, with 1-2 mg of actinide released in these conditions.

  16. Radiochemical determination of rare Earth elements in proton-irradiated lead-bismuth eutectic.

    PubMed

    Hammer, Bernadette; Neuhausen, Jörg; Boutellier, Viktor; Wohlmuther, Michael; Türler, Andreas; Schumann, Dorothea

    2015-06-01

    Various types of proton-irradiated lead-bismuth eutectic (LBE) samples from the MEGAPIE prototype spallation target were analyzed concerning their content of (148)Gd, (173)Lu, and (146)Pm by use of α- and γ-spectrometry. A radiochemical separation procedure was developed to isolate the lanthanide fraction and to prepare thin samples for α-ray measurement. The results prove a substantial depletion of these three elements in bulk samples, whereas accumulation on the LBE/steel-interfaces was observed. The amount of material accumulated on surfaces was roughly estimated by relating the values measured on the sample surfaces to the total surface of the inner target walls. The amount of (148)Gd, (173)Lu, and (146)Pm was then quantified by summing up the contributions from every sample type. The results show a reasonable agreement with theoretical predictions. The obtained results are of utmost importance for the evaluation of the performance of high-power spallation targets, especially concerning the residual nuclide production, the physicochemical behavior of the produced radionuclides during operation, and in terms of an intermediate or final disposal. PMID:25938905

  17. Optical properties of Lead bismuth borate glasses doped with neodymium oxide.

    PubMed

    Farouk, M; Abd El-Maboud, A; Ibrahim, M; Ratep, A; Kashif, I

    2015-10-01

    Neodymium doped Lead bismuth borate glasses with the composition of 25PbO-25Bi2O3-50B2O3:xNd2O3, where x=0.5, 1, 1.5 and 2 mol%, have been prepared by melt quenching technique. The behavior of the density and molar volume allows concluding that, addition of Nd2O3 leads to the formation of non-bridging oxygen. Rare earth ion parameters have been calculated and studied. The optical band gap (Eg), and band tails (Ee) were determined. Judd-Ofelt theory for the intensity analysis of induced electric dipole transitions has been applied to the measured oscillator strengths of the absorption bands to determine the three phenomenological intensity parameters Ω2, Ω4 and Ω6 for glass. It was observed that the deviation parameters, rms, was found to be 0.56:0.58(×10(-6)). The estimated Judd-Ofelt parameters were found to be Nd2O3concentration dependent. The hypersensitive transition, (4)I9/2→(4)G5/2+(2)G7/2, is closely related to Ω2 parameter. PMID:25965518

  18. Corrosion behaviour of martensitic and austenitic steels in flowing lead-bismuth eutectic

    NASA Astrophysics Data System (ADS)

    Martín-Muñoz, F. J.; Soler-Crespo, L.; Gómez-Briceño, D.

    2011-09-01

    The LINCE loop is a forced convection loop designed for long-term corrosion tests in lead-bismuth eutectic (LBE) at CIEMAT. The LBE volume of in the loop is 250 l and the maximum flow velocity in the region of specimens is approximately 1 m s -1. An oxygen control system has been implemented in the loop. The corrosion behaviour of AISI 316L and T91 steels was investigated in flowing LBE at temperatures of 575 and 725 K for exposure times of 2000, 5000 and 10,000 h. At 575 K, the results showed a good response, with no weight loss detected in any of the materials after exposure to the flowing LBE up to 10,000 h. A similar behaviour was observed for the specimens tested at 725 K during 2000 and 10,000 h. Specimens extracted at intermediate time (5000 h) showed an anomalous behaviour with important weight loss. These specimens were placed at the bottom of the hot test section, and this position probably made them to suffer an accused process of cavitation-erosion.

  19. Corrosion behaviors of US steels in flowing lead bismuth eutectic (LBE)

    NASA Astrophysics Data System (ADS)

    Zhang, Jinsuo; Li, Ning; Chen, Yitung; Rusanov, A. E.

    2005-01-01

    Corrosion tests of several US martensitic and austenitic steels were performed in a forced circulation lead-bismuth eutectic non-isothermal loop at the Institute of Physics and Power Engineering (IPPE), Russia. Tube and rod specimens of austenitic steels 316/316L, D-9, and martensitic steels HT-9, T-410 were inserted in the loop. Experiments were carried out simultaneously at 460 °C and 550 °C for 1000, 2000 and 3000 h. The flow velocity at the test sections was 1.9 m/s and the oxygen concentration in LBE was in the range of 0.03-0.05 wppm. The results showed that at 460 °C, all the test steels have satisfactory corrosion resistance: a thin protective oxide layer formed on the steel surfaces and no observable dissolution of steel components occurred. At 550 °C, rod specimens suffered rather severe local liquid metal corrosion and slot corrosion; while tube specimens were subject to oxidation and formed double-layer oxide films that can be roughly described as a porous Fe 3O 4 outer layer over a chrome-rich spinel inner layer. Neglecting the mass transfer corrosion effects by the flowing LBE, calculations based on Wagner's theory reproduce the experimental results on the oxide thickness, indicating that the oxide growth mechanism of steels in LBE is similar to that of steels in air/steam, with slight modification by dissolution and oxide dissociation at the liquid metal interface.

  20. Overview of recent studies related to lead-alloy-cooled fast reactors

    NASA Astrophysics Data System (ADS)

    Takahashi, Minoru; Sa, Rongyuan; Pramutadi, Asril; Yamaki-Irisawa, Eriko

    2012-06-01

    The recent progress of the studies related to lead alloy-cooled reactors (LFR) and the accelerator driven system (ADS) is summarized. The compatibility of materials with lead alloys has been clarified under steady and transient temperature conditions. Higher Cr content, Si and Al addition and Al-Fe alloy-coating improved the corrosion resistance of steels. The Al-Fe alloy-coated steel was not corroded even under high temperature transient conditions. The ceramics of SiC and Si3N4 are expected to be used as cladding material for high temperature LFR. For the analytical consideration of mass transport in lead alloy circuit, the diffusion coefficient of Ni was measured using the capillary methods. A new bubble visualization method in LBE using gamma-ray radiography was developed. The thermal interaction of lead-bismuth eutectic (LBE) and lead droplets with sub-cooled water, and the behaviors of droplet fragmentation were investigated, and the visualization of volatile liquids in high temperature liquids was achieved.

  1. Fuel Development For Gas-Cooled Fast Reactors

    SciTech Connect

    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.

  2. Fuel development for gas-cooled fast reactors

    NASA Astrophysics Data System (ADS)

    Meyer, M. K.; Fielding, R.; Gan, J.

    2007-09-01

    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 [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. 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 dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US 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.

  3. Fast neutron dosimeter using Cooled Optically Stimulated Luminescence (COSL)

    SciTech Connect

    Eschbach, P.A.; Miller, S.D.

    1991-10-01

    Data is presented that demonstrates the concept of a fast neutron dosimeter using Cooled Optically Stimulated Luminescence. CaF{sub 2}:Mn powder, compounded with polyethylene, was injection molded and pressed into 0.1-cm-thick sheets. The sheets were then cut to form dosimeters with dimensions, 1.25 cm by 1.25 cm. After a laser anneal, the dosimeters were exposed to various amounts (from 10 mSv to 100 mSv) of fast {sup 252}Cf neutrons. The exposed dosimeters were cooled to liquid nitrogen temperature, stimulated with laser light, and then allowed to warm up to room temperature whereupon the dose dependent luminescence was recorded with a photon counting system. When the control and gamma components were subtracted from the {sup 252}Cf response, a dose-dependent neutron response was observed. The design, construction, and preliminary performance of an automated system for the dose interrogation of individual CaF{sub 2}:Mn grains within the polyethylene matrix will also be discussed. The system uses a small CO{sub 2} laser to heat areas of the cooled dosimeter to room temperature. If the readout of very small grain within the plastic matrix is successful, it will enhance the neutron to gamma response of the dosimeter.

  4. High performance infrared fast cooled detectors for missile applications

    NASA Astrophysics Data System (ADS)

    Reibel, Yann; Espuno, Laurent; Taalat, Rachid; Sultan, Ahmad; Cassaigne, Pierre; Matallah, Noura

    2016-05-01

    SOFRADIR was selected in the late 90's for the production of 320×256 MW detectors for major European missile programs. This experience has established our company as a key player in the field of missile programs. SOFRADIR has since developed a vast portfolio of lightweight, compact and high performance JT-based solutions for missiles. ALTAN is a 384x288 Mid Wave infrared detector with 15μm pixel pitch, and is offered in a miniature ultra-fast Joule- Thomson cooled Dewar. Since Sofradir offers both Indium Antimonide (InSb) and Mercury Cadmium Telluride technologies (MCT), we are able to deliver the detectors best suited to customers' needs. In this paper we are discussing different figures of merit for very compact and innovative JT-cooled detectors and are highlighting the challenges for infrared detection technologies.

  5. Decay heat removal in GEN IV gas cooled fast reactors.

    SciTech Connect

    Cheng, L. Y.; Wei, T. Y. C.

    2009-08-01

    The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

  6. Shape optimization of a sodium cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Schmitt, Damien; Allaire, Grégoire; Pantz, Olivier; Pozin, Nicolas

    2014-06-01

    Traditional designs of sodium cooled fast reactors have a positive sodium expansion feedback. During a loss of flow transient without scram, sodium heating and boiling thus insert a positive reactivity and prevents the power from decreasing. Recent studies led at CEA, AREVA and EDF show that cores with complex geometries can feature a very low or even a negative sodium void worth.(1, 2) Usual optimization methods for core conception are based on a parametric description of a given core design(3).(4) New core concepts and shapes can then only be found by hand. Shape optimization methods have proven very efficient in the conception of optimal structures under thermal or mechanical constraints.(5, 6) First studies show that these methods could be applied to sodium cooled core conception.(7) In this paper, a shape optimization method is applied to the conception of a sodium cooled fast reactor core with low sodium void worth. An objective function to be minimized is defined. It includes the reactivity change induced by a 1% sodium density decrease. The optimization variable is a displacement field changing the core geometry from one shape to another. Additionally, a parametric optimization of the plutonium content distribution of the core is made, so as to ensure that the core is kept critical, and that the power shape is flat enough. The final shape obtained must then be adjusted to a get realistic core layout. Its caracteristics can be checked with reference neutronic codes such as ERANOS. Thanks to this method, new shapes of reactor cores could be inferred, and lead to new design ideas.

  7. Corrosion-erosion test of SS316L grain boundary engineering material (GBEM) in lead bismuth flowing loop

    NASA Astrophysics Data System (ADS)

    Saito, Shigeru; Kikuchi, Kenji; Hamaguchi, Dai; Tezuka, Masao; Miyagi, Masanori; Kokawa, Hiroyuki; Watanabe, Seiichi

    2012-12-01

    To evaluate the lifetime of structural materials utilized in a spallation neutron source, corrosion tests in lead-bismuth eutectic (LBE) have been done at JAEA. Austenitic steels are preferable as the structural material for ADS. However, previous studies have revealed that austenitic steel SS316 shows severe corrosion-erosion in LBE because of LBE penetration through grain boundaries and separation of grains. So it was considered that GBE (grain-boundary engineered) materials may be effective to improve the corrosion resistance of austenitic steels in LBE. In this study, the results of corrosion tests on austenitic steel SS316L-BM (base metal) and SS316L-GBEM (grain-boundary-engineered material) under flowing LBE conditions will be reported. The corrosion test was performed using the JAEA lead-bismuth material corrosion loop (JLBL-1). The experimental conditions were as follows: The high and low temperature parts of the loop were 450 °C and 350 °C, respectively. The flow velocity at the test specimens was about 0.7 m/s. The oxygen concentration in LBE was not controlled and was estimated to have been very low. After the 3600 h of operation, macroscopic, SEM, and SIM observations and EDX analysis were carried out. The results showed that the corrosion depth and LBE penetration through the grain boundaries of the 316SS-GBEM were smaller than those of the 316SS-BM.

  8. Fuels for sodium-cooled fast reactors: US perspective

    NASA Astrophysics Data System (ADS)

    Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.

    2007-09-01

    The US experience with mixed oxide, metal, and mixed carbide fuels is substantial, comprised of irradiation of over 50 000 MOX rods, over 130 000 metal rods, and 600 mixed carbide rods, in EBR-II and FFTF alone. All three types have been demonstrated capable of fuel utilization at or above 200 GWd/MTHM. To varying degrees, life-limiting phenomena for each type have been identified and investigated, and there are no disqualifying safety-related fuel behaviors. All three fuel types appear capable of meeting requirements of sodium-cooled fast reactor fuels, with reliability of mixed oxide and metal fuel well established. Improvements in irradiation performance of cladding and duct alloys have been a key development in moving these fuel designs toward higher-burnup potential. Selection of one fuel system over another will depend on circumstances particular to the application and on issues other than fuel performance, such as fabrication cost or overall system safety performance.

  9. Gas-Cooled Fast Reactor (GFR) Decay Heat Removal Concepts

    SciTech Connect

    K. D. Weaver; L-Y. Cheng; H. Ludewig; J. Jo

    2005-09-01

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with an outlet temperature of 850ºC at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report is a compilation of work performed on decay heat removal systems for a 2400 MWt GFR during this fiscal year (FY05).

  10. Fast radiative cooling of anthracene: Dependence on internal energy

    NASA Astrophysics Data System (ADS)

    Martin, S.; Ji, M.; Bernard, J.; Brédy, R.; Concina, B.; Allouche, A. R.; Joblin, C.; Ortega, C.; Montagne, G.; Cassimi, A.; Ngono-Ravache, Y.; Chen, L.

    2015-11-01

    Fast radiative cooling of anthracene cations (C14H10 ) + is studied with a compact electrostatic storage device, the Mini-Ring. The time evolution of the internal energy distribution of the stored ions is probed in a time range from 3 to 7 ms using laser-induced dissociation with 3.49-eV photons. The population decay rate due to radiative emission is measured to vary from 25 to 450 s-1 as a function of the excitation energy in the range from 6 to 7.4 eV. After corrections of the infrared emission effect via vibrational transitions, the fluorescence emission rate due to electronic transitions from thermally excited electronic states is estimated and compared with a statistical molecular approach. In the considered internal energy range, the radiative cooling process is found to be dominated by the electronic transition, in good agreement with our previous work [S. Martin et al., Phys. Rev. Lett. 110, 063003 (2013), 10.1103/PhysRevLett.110.063003] focused on a narrower energy range.

  11. Experimental Study on Flow Technology and Steel Corrosion of Lead-Bismuth

    SciTech Connect

    Minoru Takahashi; Hiroshi Sekimoto; Kotaro Ishikawa; Naoki Sawada; Tadashi Suzuki; Susumu Yoshida; Toyohiko Yano; Masamitsu Imai; Koji Hata; Suizheng Qiu

    2002-07-01

    For the feasibility study of Pb-Bi-cooled fast reactors (FR) and the Pb-Bi target of accelerator-driven nuclear transmutation systems, Pb-Bi flow technologies were developed and steel corrosion behavior in a Pb-Bi flow was investigated using a Pb-Bi circulation loop. The performance of an electro-magnetic flow meter with electrically insulated electrodes plated with Rh was better than those of conventional and tubular types. Oxygen concentration was controlled by continuous injection of Ar, H{sub 2} and H{sub 2}O mixture gas into the Pb-Bi flow. In order to have desired oxygen potential, the partial pressure ratio of P{sub H{sub 2}}/P{sub H{sub 2}}{sub O} was chosen in the range from 0.12 to 2.2 by bubbling the mixture of Ar and H{sub 2} in water columns at the room temperature. By injecting the mixture gas into the loop for sufficient time, the oxygen potentials measured by the oxygen sensor made of solid electrolyte ZrO{sub 2}-Y{sub 2}O{sub 3} agreed well with those in the injected gas mixture. In the first corrosion test, steels were exposed to a Pb-Bi flow at the temperature of 550 deg. C, the velocity of 2 m/s and the oxygen concentration of {approx}5.0x10{sup -7} wt.% for 959 hours. It was found that the weight loss was larger in the order of SS316, low Cr steel (SCM420) and high Cr steels (STBA26, SUS405, SUS430). Corrosion was suppressed by a Cr oxide layer for high Cr steels. A porous layer was formed on SS316 surface due to high solubility of Ni in Pb-Bi,. In the second corrosion test, the oxygen concentration was kept at 3.6x10{sup -7} wt.% by injecting Ar, H{sub 2} and H{sub 2}O mixture gas into a Pb-Bi flow, and steels were exposed to a Pb-Bi flow at the temperature of 550 deg. C, the velocity of 2 m/s for 1000 hours. Serious erosion damage was observed in SCM420 at the entrance, and some erosion damages appeared in low Cr steels: SCM420, F82H, STBA26 and HCM12 downstream. Crack type damage was observed on the surface of HCM12, and pitting-type damage

  12. Development of a plant dynamics computer code for analysis of a supercritical carbon dioxide Brayton cycle energy converter coupled to a natural circulation lead-cooled fast reactor.

    SciTech Connect

    Moisseytsev, A.; Sienicki, J. J.

    2007-03-08

    STAR-LM is a lead-cooled pool-type fast reactor concept operating under natural circulation of the coolant. The reactor core power is 400 MWt. The open-lattice core consists of fuel pins attached to the core support plate, (the does not consist of removable fuel assemblies). The coolant flows outside of the fuel pins. The fuel is transuranic nitride, fabricated from reprocessed LWR spent fuel. The cladding material is HT-9 stainless steel; the steady-state peak cladding temperature is 650 C. The coolant is single-phase liquid lead under atmospheric pressure; the core inlet and outlet temperatures are 438 C and 578 C, respectively. (The Pb coolant freezing and boiling temperatures are 327 C and 1749 C, respectively). The coolant is contained inside of a reactor vessel. The vessel material is Type 316 stainless steel. The reactor is autonomous meaning that the reactor power is self-regulated based on inherent reactivity feedbacks and no external power control (through control rods) is utilized. The shutdown (scram) control rods are used for startup and shutdown and to stop the fission reaction in case of an emergency. The heat from the reactor is transferred to the S-CO{sub 2} Brayton cycle in in-reactor heat exchangers (IRHX) located inside the reactor vessel. The IRHXs are shell-and-tube type heat exchangers with lead flowing downwards on the shell side and CO{sub 2} flowing upwards on the tube side. No intermediate circuit is utilized. The guard vessel surrounds the reactor vessel to contain the coolant, in the very unlikely event of reactor vessel failure. The Reactor Vessel Auxiliary Cooling System (RVACS) implementing the natural circulation of air flowing upwards over the guard vessel is used to cool the reactor, in the case of loss of normal heat removal through the IRHXs. The RVACS is always in operation. The gap between the vessels is filled with liquid lead-bismuth eutectic (LBE) to enhance the heat removal by air by significantly reducing the thermal

  13. Post-irradiation analysis of an ISOLDE lead-bismuth target: Stable and long-lived noble gas nuclides

    NASA Astrophysics Data System (ADS)

    Leya, I.; Grimberg, A.; David, J.-C.; Schumann, D.; Neuhausen, J.; Zanini, L.; Noah, E.

    2016-07-01

    We measured the isotopic concentrations of long-lived and stable He, Ne, Ar, Kr, and Xe isotopes in a sample from a lead-bismuth eutectic target irradiated with 1.0 and 1.4 GeV protons. Our data indicate for most noble gases nearly complete release with retention fractions in the range of percent or less. Higher retention fractions result from the decay of long-lived radioactive progenitors from groups 1, 2, or 7 of the periodic table. From the data we can calculate a retention fraction for 3H of 2-3%. For alkaline metals we find retention fractions of about 10%, 30%, and 50% for Na, Rb, and Cs, respectively. For the alkaline earth metal Ba we found complete retention. Finally, the measured Kr and Xe concentrations indicate that there was some release of the halogens Br and I during and/or after the irradiation.

  14. Development of tellurium oxide and lead-bismuth oxide glasses for mid-wave infra-red transmission optics

    NASA Astrophysics Data System (ADS)

    Zhou, Beiming; Rapp, Charles F.; Driver, John K.; Myers, Michael J.; Myers, John D.; Goldstein, Jonathan; Utano, Rich; Gupta, Shantanu

    2013-03-01

    Heavy metal oxide glasses exhibiting high transmission in the Mid-Wave Infra-Red (MWIR) spectrum are often difficult to manufacture in large sizes with optimized physical and optical properties. In this work, we researched and developed improved tellurium-zinc-barium and lead-bismuth-gallium heavy metal oxide glasses for use in the manufacture of fiber optics, optical components and laser gain materials. Two glass families were investigated, one based upon tellurium and another based on lead-bismuth. Glass compositions were optimized for stability and high transmission in the MWIR. Targeted glass specifications included low hydroxyl concentration, extended MWIR transmission window, and high resistance against devitrification upon heating. Work included the processing of high purity raw materials, melting under controlled dry Redox balanced atmosphere, finning, casting and annealing. Batch melts as large as 4 kilograms were sprue cast into aluminum and stainless steel molds or temperature controlled bronze tube with mechanical bait. Small (100g) test melts were typically processed in-situ in a 5%Au°/95%Pt° crucible. Our group manufactured and evaluated over 100 different experimental heavy metal glass compositions during a two year period. A wide range of glass melting, fining, casting techniques and experimental protocols were employed. MWIR glass applications include remote sensing, directional infrared counter measures, detection of explosives and chemical warfare agents, laser detection tracking and ranging, range gated imaging and spectroscopy. Enhanced long range mid-infrared sensor performance is optimized when operating in the atmospheric windows from ~ 2.0 to 2.4μm, ~ 3.5 to 4.3μm and ~ 4.5 to 5.0μm.

  15. Gas-Cooled Fast Reactor (GFR) FY04 Annual Report

    SciTech Connect

    K. D. Weaver; T. C. Totemeier; D. E. Clark; E. E. Feldman; E. A. Hoffman; R. B. Vilim; T. Y. C. Wei; J. Gan; M. K. Meyer; W. F. Gale; M. J. Driscoll; M. Golay; G. Apostolakis; K. Czerwinski

    2004-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  16. Gas-Cooled Fast Reactor (GFR) FY05 Annual Report

    SciTech Connect

    K. D. Weaver; T. Marshall; T. Totemeier; J. Gan; E.E. Feldman; E.A Hoffman; R.F. Kulak; I.U. Therios; C. P. Tzanos; T.Y.C. Wei; L-Y. Cheng; H. Ludewig; J. Jo; R. Nanstad; W. Corwin; V. G. Krishnardula; W. F. Gale; J. W. Fergus; P. Sabharwall; T. Allen

    2005-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in

  17. Fast cooling in dispersively and dissipatively coupled optomechanics.

    PubMed

    Chen, Tian; Wang, Xiang-Bin

    2015-01-01

    The cooling performance of an optomechanical system comprising both dispersive and dissipative coupling is studied. Here, we present a scheme to cool a mechanical resonator to its ground state in finite time using a chirped pulse. We show that there is distinct advantage in using the chirp-pulse scheme to cool a resonator rapidly. The cooling behaviors of dispersively and dissipatively coupled system is also explored with different types of incident pulses and different coupling strengths. Our scheme is feasible in cooling the resonator for a wide range of the parameter region. PMID:25582660

  18. A Comparison of Long-Lived, Prolieration Resistant Fast Reactors

    SciTech Connect

    Weaver, Kevan Dean; Herring, James Stephen; Mac Donald, Philip Elsworth

    2001-09-01

    Nuclear power is expected to play a significant role in meeting future electricity needs, and in significantly reducing emissions compared to fossil-fueled power plants. However, the next generation of nuclear power plants will be expected to demonstrate significant advancements in economics, safety, waste disposal, and proliferation resistance. Many reactor types have been proposed for “Generation IV”, some of which have been fast reactors. The work discussed in here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal of the entire project is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. The goal of the work presented in this paper is to investigate and compare a variety of possible fuel types, looking for optimum economics for an actinide burning, low cost of electricity, reactor design using sodium or lead-bismuth as the coolant.

  19. Optical absorption and fluorescence properties of Er{sup 3+}/Yb{sup 3+} codoped lead bismuth alumina borate glasses

    SciTech Connect

    Goud, K. Krishna Murthy Reddy, M. Chandra Shekhar Rao, B. Appa

    2014-04-24

    Lead bismuth alumina borate glasses codoped with Er{sup 3+}/Yb{sup 3+} were prepared by melt quenching technique. Optical absorption, FTIR and photoluminescence spectra of these glasses have been studied. Judd-Ofelt theory has been applied to to the f ↔ f transitions for evaluating Ω{sub 2}, Ω{sub 4} and Ω{sub 6} parameters. Radiative properties like branching ratio β{sub r} and the radiative life time τ{sub R} have been determined on the basis of Judd-Ofelt theory. Upconversion emissions have been observed under 980nm laser excitation at room temperature. Green and red up-conversion emissions are centered at 530, 550 and 656 nm corresponding to {sup 2}H{sub 11/2}→{sup 4}I{sub 15/2}, {sup 4}S{sub 3/2}→{sup 4}I{sub 15/2} and {sup 4}F{sub 9/2}→{sup 4}I{sub 15/2} transitions of Er{sup 3+} respectively. The results obtained are discussed quantitatively based on the energy transfer between Yb{sup 3+} and Er{sup 3+}.

  20. Stress corrosion behavior of T91 steel in static lead-bismuth eutectic at 480 °C

    NASA Astrophysics Data System (ADS)

    Liu, Jing; Jiang, Zhizhong; Tian, Shujian; Huang, Qunying; Liu, Yuejing

    2016-01-01

    The corrosion behavior of stressed C-rings made of martensitic steel T91 was investigated through constant strain tests. The specimens with different initial hoop stresses (0 MPa, 150 MPa and 300 MPa) were exposed to static oxygen saturated lead-bismuth eutectic (LBE) at 480 °C for 500 h, 1000 h and 1500 h, respectively. The results showed that no crack was found on the outer surface of all the specimens after exposure; and the microscopic analysis showed that the specimens were covered with two oxide layers, which included a magnetite outer layer and a Fe-Cr spinel inner layer. The transformation of spinel into magnetite at the spinel/magnetite interface might be promoted by stress, which increased the difference between the thickness of the inner and outer layers. Moreover, the steel loss was estimated by the observed oxide layers; it increased rapidly when the stress was above 300 MPa, and was about 1.3 times of when the stress was absent.

  1. Characterization of the mechanism of bi-layer oxide growth on austenitic stainless steels 316L and D9 in oxygen-controlled Lead-Bismuth Eutectic (LBE)

    NASA Astrophysics Data System (ADS)

    Koury, Daniel

    Lead Bismuth Eutectic (LBE) has been proposed for use in programs for accelerator-based and reactor-based transmutation of nuclear waste. LBE is a leading candidate material as a spallation target (in accelerator-based transmutation) and an option for the sub-critical blanket coolant. The corrosion by LBE of annealed and cold-rolled 316L stainless steels, and the modified austenitic stainless steel alloy D9, has been studied using Scanning Electron Microscopy (SEM), Electron Probe Micro Analysis (EPMA), and X-ray Photoelectron Spectroscopy (XPS). Exposed and unexposed samples have been compared and the differences studied. Small amounts of surface contamination are present on the samples and have been removed by ion-beam sputtering. The unexposed samples reveal typical stainless steel characteristics: a chromium oxide passivation surface layer and metallic iron and nickel. The exposed samples show protective iron oxide and chromium oxide growths on the surface. Oxygen takes many forms on the exposed samples, including oxides of iron and chromium, carbonates, and organic acids from subsequent handling after exposure to LBE. Different types of surface preparation have lead to considerably different modes of corrosion. The cold-rold samples were resistant to thick oxide growth, having only a thin (< 1 mum), dense chromium-rich oxide. The annealed 316L and D9 samples developed thick, bi-layered oxides, the inner layer consisting of chromium-rich oxides (likely spinel) and the outer layer consisting mostly of iron oxides. The cold-rolled samples were able to maintain a thin chromium oxide layer because of the surface work performed on it, as ample diffusion pathways provided an adequate supply of chromium atoms. The annealed samples grew thick oxides because iron was the primary diffusant, as there are fewer fast-diffusion pathways and therefore an amount of chromium insufficient to maintain a chromium based oxide. Even the thick oxide, however, can prolong the life of

  2. Compatibility of ferritic-martensitic steel T91 welds with liquid lead-bismuth eutectic: Comparison between TIG and EB welds

    NASA Astrophysics Data System (ADS)

    Van den Bosch, J.; Coen, G.; Van Renterghem, W.; Almazouzi, A.

    2010-01-01

    The 9 wt.% chromium ferritic-martensitic steel T91 is being considered as candidate structural material for a future experimental accelerator driven system (XT-ADS). This material and its welded connections would need to be used in contact with liquid lead-bismuth eutectic (LBE), under high irradiation doses. Both unirradiated tungsten inert gas (TIG) and electron beam (EB) welds of T91 have been examined by means of metallography, scanning electron microscopy (SEM-EDX), transmission electron microscopy (TEM), Vickers hardness measurements and tensile testing in both gas and liquid lead-bismuth environment. The TIG weld was commercially produced and post weld heat treated by a certified welding company while the post weld heat treatment of the experimental EB weld was optimized in terms of the Vickers hardness profile across the welded joint. The mechanical properties of the T91 TIG and EB welds in contact with LBE have been examined using slow strain rate tensile testing (SSRT) in LBE at 350 °C. All welds showed good mechanical behaviour in gas environment but total elongation was strongly reduced due to liquid metal embrittlement (LME) when tested in liquid lead-bismuth eutectic environment. The reduction in total elongation due to LME was larger for the commercially TIG welded joint than for the EB welded joint.

  3. Fast Quasi-Adiabatic Gas Cooling: An Experiment Revisited

    ERIC Educational Resources Information Center

    Oss, S.; Gratton, L. M.; Calza, G.; Lopez-Arias, T.

    2012-01-01

    The well-known experiment of the rapid expansion and cooling of the air contained in a bottle is performed with a rapidly responsive, yet very cheap thermometer. The adiabatic, low temperature limit is approached quite closely and measured with our apparatus. A straightforward theoretical model for this process is also presented and discussed.…

  4. Fast Cooling and Vitrification of Aqueous Solutions for Cryopreservation

    NASA Astrophysics Data System (ADS)

    Warkentin, Matt; Husseini, Naji; Berejnov, Viatcheslav; Thorne, Robert

    2006-03-01

    In many applications, a small volume of aqueous solution must be cooled at a rate sufficient to produce amorphous solid water. Two prominent examples include flash-freezing of protein crystals for X-ray data collection and freezing of cells (i.e. spermatozoa) for cryopreservation. The cooling rate required to vitrify pure water (˜10^6 K/s) is unattainable for volumes that might contain cells or protein crystals, but the required rate can be reduced by adding cryoprotectants. We report the first measurements of the critical concentration required to produce a vitrified sample as a function of the sample's volume, the cryogen into which the sample is plunged, and the temperature of the cryogen, for a wide range of cryoprotectants. These experiments have broad practical consequences for cryopreservation, and provide insight into the physics of glass formation in aqueous systems.

  5. Fast optical cooling of a nanomechanical cantilever by a dynamical Stark-shift gate

    PubMed Central

    Yan, Leilei; Zhang, Jian-Qi; Zhang, Shuo; Feng, Mang

    2015-01-01

    The efficient cooling of nanomechanical resonators is essential to exploration of quantum properties of the macroscopic or mesoscopic systems. We propose such a laser-cooling scheme for a nanomechanical cantilever, which works even for the low-frequency mechanical mode and under weak cooling lasers. The cantilever is coupled by a diamond nitrogen-vacancy center under a strong magnetic field gradient and the cooling is assisted by a dynamical Stark-shift gate. Our scheme can effectively enhance the desired cooling efficiency by avoiding the off-resonant and undesired carrier transitions, and thereby cool the cantilever down to the vicinity of the vibrational ground state in a fast fashion. PMID:26455901

  6. Low cryoprotectant concentrations and fast cooling for nematode cryostorage.

    PubMed

    Irdani, Tiziana; Scotto, Cristina; Roversi, Pio Federico

    2011-08-01

    Cryopreservation protocols based on slow freezing or vitrification often result in cell injury due to ice formation, cell dehydration and/or toxic concentrations of cryoprotectant (CPA). In this study, we present a cryopreservation technique based on low, non-toxic concentrations of cryoprotectants (≈ 2-4M) combined with a rapid cooling rate in the liquid nitrogen phase (-196°C). Protocols for successfully cryopreserving the plant parasitic nematodes Globodera tabacum tabacum, Heterodera schachtii and Meloidogyne incognita were developed, as demonstrated by the high survival rates and reproducibility of cyst and root-knot nematode species post-cryostorage. This approach for effective cryopreservation of viable plant-parasitic nematodes was developed by inducing an "apparent vitrification" by rapid cooling of the microscopic samples in less than 2M of cryoprotectant. The extremely thin structure (15-20 μm width, 350-400 μm length) of these nematodes, in combination with a direct and rapid exposure to LN(2), likely prevents the formation of damaging ice crystals. Moreover, this procedure results in viability of both short- and long-cryostorage samples. These techniques could potentially be used for the near-indefinite preservation of thousands of different nematode species. A cryo-nematode collection produced in our lab is available and presented here. PMID:21524646

  7. Efficient, Indirect Transverse Laser Cooling of a Fast Stored Ion Beam

    SciTech Connect

    Miesner, H.; Grimm, R.; Grieser, M.; Habs, D.; Schwalm, D.; Wanner, B.; Wolf, A.

    1996-07-01

    Three-dimensional laser cooling of a fast stored ion beam has been demonstrated at the Heidelberg Test Storage Ring. With a purely longitudinal cooling force applied to a 7.3 MeV {sup 3}Be{sup +} beam, we have observed an efficient transverse cooling effect. We interpret this observation as being due to a thermal intrabeam relaxation between the different degrees of freedom that is caused by Coulomb collisions of the stored particles. {copyright} {ital 1996 The American Physical Society.}

  8. Microstructure and Mechanism of Strengthening of Microalloyed Pipeline Steel: Ultra-Fast Cooling (UFC) Versus Laminar Cooling (LC)

    NASA Astrophysics Data System (ADS)

    Zhao, J.; Wang, X.; Hu, W.; Kang, J.; Yuan, G.; Di, H.; Misra, R. D. K.

    2016-05-01

    A novel thermo-mechanical controlled processing (TMCP) schedule involving ultra-fast cooling (UFC) technique was used to process X70 (420 MPa) microalloyed pipeline steel with high strength-high toughness combination. A relative comparison is made between microstructure and mechanical properties between conventionally processed (CP) and ultra-fast cooled (UFC) pipeline steels, together with differences in strengthening mechanisms with respect to both types of processes. UFC-processed steel exhibited best combination of strength and good toughness compared to the CP process. The microstructure of CP pipeline steel mainly consisted of acicular ferrite (AF), bainitic ferrite (BF), and dispersed secondary martensite/austenite (M/A) constituent and a small fraction of fine quasi-polygonal ferrite. In contrast, the microstructure of UFC-processed pipeline steel was predominantly composed of finer AF, BF, and dispersed M/A constituent. The primary strengthening mechanisms in UFC pipeline steel were grain size strengthening and dislocation strengthening with strength increment of ~277 and ~151 MPa, respectively. However, the strengthening contribution in CP steel was related to grain size strengthening, dislocation strengthening, and precipitation strengthening, and the corresponding strength increments were ~212, ~149 and ~86 MPa, respectively. The decrease in strength induced by reducing Nb and Cr in UFC pipeline steel was compensated by enhancing the contribution of grain size strengthening in the UFC process. In conclusion, cooling schedule of UFC combined with LC is a promising method for processing low-cost pipeline steels.

  9. Microstructure and Mechanism of Strengthening of Microalloyed Pipeline Steel: Ultra-Fast Cooling (UFC) Versus Laminar Cooling (LC)

    NASA Astrophysics Data System (ADS)

    Zhao, J.; Wang, X.; Hu, W.; Kang, J.; Yuan, G.; Di, H.; Misra, R. D. K.

    2016-06-01

    A novel thermo-mechanical controlled processing (TMCP) schedule involving ultra-fast cooling (UFC) technique was used to process X70 (420 MPa) microalloyed pipeline steel with high strength-high toughness combination. A relative comparison is made between microstructure and mechanical properties between conventionally processed (CP) and ultra-fast cooled (UFC) pipeline steels, together with differences in strengthening mechanisms with respect to both types of processes. UFC-processed steel exhibited best combination of strength and good toughness compared to the CP process. The microstructure of CP pipeline steel mainly consisted of acicular ferrite (AF), bainitic ferrite (BF), and dispersed secondary martensite/austenite (M/A) constituent and a small fraction of fine quasi-polygonal ferrite. In contrast, the microstructure of UFC-processed pipeline steel was predominantly composed of finer AF, BF, and dispersed M/A constituent. The primary strengthening mechanisms in UFC pipeline steel were grain size strengthening and dislocation strengthening with strength increment of ~277 and ~151 MPa, respectively. However, the strengthening contribution in CP steel was related to grain size strengthening, dislocation strengthening, and precipitation strengthening, and the corresponding strength increments were ~212, ~149 and ~86 MPa, respectively. The decrease in strength induced by reducing Nb and Cr in UFC pipeline steel was compensated by enhancing the contribution of grain size strengthening in the UFC process. In conclusion, cooling schedule of UFC combined with LC is a promising method for processing low-cost pipeline steels.

  10. Fast cooling for a system of stochastic oscillators

    NASA Astrophysics Data System (ADS)

    Chen, Yongxin; Georgiou, Tryphon T.; Pavon, Michele

    2015-11-01

    We study feedback control of coupled nonlinear stochastic oscillators in a force field. We first consider the problem of asymptotically driving the system to a desired steady state corresponding to reduced thermal noise. Among the feedback controls achieving the desired asymptotic transfer, we find that the most efficient one from an energy point of view is characterized by time-reversibility. We also extend the theory of Schrödinger bridges to this model, thereby steering the system in finite time and with minimum effort to a target steady-state distribution. The system can then be maintained in this state through the optimal steady-state feedback control. The solution, in the finite-horizon case, involves a space-time harmonic function φ, and -logφ plays the role of an artificial, time-varying potential in which the desired evolution occurs. This framework appears extremely general and flexible and can be viewed as a considerable generalization of existing active control strategies such as macromolecular cooling. In the case of a quadratic potential, the results assume a form particularly attractive from the algorithmic viewpoint as the optimal control can be computed via deterministic matricial differential equations. An example involving inertial particles illustrates both transient and steady state optimal feedback control.

  11. Fast cooling for a system of stochastic oscillators

    SciTech Connect

    Chen, Yongxin Georgiou, Tryphon T.; Pavon, Michele

    2015-11-15

    We study feedback control of coupled nonlinear stochastic oscillators in a force field. We first consider the problem of asymptotically driving the system to a desired steady state corresponding to reduced thermal noise. Among the feedback controls achieving the desired asymptotic transfer, we find that the most efficient one from an energy point of view is characterized by time-reversibility. We also extend the theory of Schrödinger bridges to this model, thereby steering the system in finite time and with minimum effort to a target steady-state distribution. The system can then be maintained in this state through the optimal steady-state feedback control. The solution, in the finite-horizon case, involves a space-time harmonic function φ, and −logφ plays the role of an artificial, time-varying potential in which the desired evolution occurs. This framework appears extremely general and flexible and can be viewed as a considerable generalization of existing active control strategies such as macromolecular cooling. In the case of a quadratic potential, the results assume a form particularly attractive from the algorithmic viewpoint as the optimal control can be computed via deterministic matricial differential equations. An example involving inertial particles illustrates both transient and steady state optimal feedback control.

  12. Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

    SciTech Connect

    Kasten, P.R.

    1981-09-01

    Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies. (DLC)

  13. Emergency cooling down of fast-neutron reactors by natural convection (a review)

    NASA Astrophysics Data System (ADS)

    Zhukov, A. V.; Sorokin, A. P.; Kuzina, Yu. A.

    2013-05-01

    Various methods for emergency cooling down of fast-neutron reactors by natural convection are discussed. The effectiveness of using natural convection for these purposes is demonstrated. The operating principles of different passive decay heat removal systems intended for cooling down a reactor are explained. Experimental investigations carried out in Russia for substantiating the removal of heat in cooling down fast-neutron reactors are described. These investigations include experimental works on studying thermal hydraulics in small-scale simulation facilities containing the characteristic components of a reactor (reactor core elements, above-core structure, immersed and intermediate heat exchangers, pumps, etc.). It is pointed out that a system that uses leaks of coolant between fuel assemblies holds promise for fast-neutron reactor cooldown purposes. Foreign investigations on this problem area are considered with making special emphasis on the RAMONA and NEPTUN water models. A conclusion is drawn about the possibility of using natural convection as the main method for passively removing heat in cooling down fast-neutron reactors, which is confirmed experimentally both in Russia and abroad.

  14. Emergency Decay Heat Removal in a GEN-IV Gas-Cooled Fast Reactor

    SciTech Connect

    Cheng, Lap Y.; Ludewig, Hans; Jo, Jae

    2006-07-01

    A series of transient analyses using the system code RELAP5-3d has been performed to confirm the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400 MWt GEN-IV gas-cooled fast reactor. The accident sequence of interest is a station blackout simultaneous with a small break (10 sq.inch/0.645 m{sup 2}) in the reactor vessel. The analyses cover the three phases of decay heat removal in a depressurization accident: (1) forced flow cooling by the power conversion unit (PCU) coast down, (2) active forced flow cooling by a battery powered blower, and (3) passive cooling by natural circulation. The blower is part of an emergency cooling system (ECS) that by design is to sustain passive decay heat removal via natural circulation cooling 24 hours after shutdown. The RELAP5 model includes the helium-cooled reactor, the ECS (primary and secondary side), the PCU with all the rotating machinery (turbine and compressors) and the heat transfer components (recuperator, pre-cooler and inter-cooler), and the guard containment that surrounds the reactor and the PCU. The transient analysis has demonstrated the effectiveness of passive decay heat removal by natural circulation cooling when the guard containment pressure is maintained at or above 800 kPa. (authors)

  15. Rapid hydrothermal cooling above the axial melt lens at fast-spreading mid-ocean ridge.

    PubMed

    Zhang, Chao; Koepke, Juergen; Kirchner, Clemens; Götze, Niko; Behrens, Harald

    2014-01-01

    Axial melt lenses sandwiched between the lower oceanic crust and the sheeted dike sequences at fast-spreading mid-ocean ridges are assumed to be the major magma source of oceanic crust accretion. According to the widely discussed "gabbro glacier" model, the formation of the lower oceanic crust requires efficient cooling of the axial melt lens, leading to partial crystallization and crystal-melt mush subsiding down to lower crust. These processes are believed to be controlled by periodical magma replenishment and hydrothermal circulation above the melt lens. Here we quantify the cooling rate above melt lens using chemical zoning of plagioclase from hornfelsic recrystallized sheeted dikes drilled from the East Pacific at the Integrated Ocean Drilling Program Hole 1256D. We estimate the cooling rate using a forward modelling approach based on CaAl-NaSi interdiffusion in plagioclase. The results show that cooling from the peak thermal overprint at 1000-1050°C to 600°C are yielded within about 10-30 years as a result of hydrothermal circulation above melt lens during magma starvation. The estimated rapid hydrothermal cooling explains how the effective heat extraction from melt lens is achieved at fast-spreading mid-ocean ridges. PMID:25209311

  16. Rapid hydrothermal cooling above the axial melt lens at fast-spreading mid-ocean ridge

    PubMed Central

    Zhang, Chao; Koepke, Juergen; Kirchner, Clemens; Götze, Niko; Behrens, Harald

    2014-01-01

    Axial melt lenses sandwiched between the lower oceanic crust and the sheeted dike sequences at fast-spreading mid-ocean ridges are assumed to be the major magma source of oceanic crust accretion. According to the widely discussed “gabbro glacier” model, the formation of the lower oceanic crust requires efficient cooling of the axial melt lens, leading to partial crystallization and crystal-melt mush subsiding down to lower crust. These processes are believed to be controlled by periodical magma replenishment and hydrothermal circulation above the melt lens. Here we quantify the cooling rate above melt lens using chemical zoning of plagioclase from hornfelsic recrystallized sheeted dikes drilled from the East Pacific at the Integrated Ocean Drilling Program Hole 1256D. We estimate the cooling rate using a forward modelling approach based on CaAl-NaSi interdiffusion in plagioclase. The results show that cooling from the peak thermal overprint at 1000–1050°C to 600°C are yielded within about 10–30 years as a result of hydrothermal circulation above melt lens during magma starvation. The estimated rapid hydrothermal cooling explains how the effective heat extraction from melt lens is achieved at fast-spreading mid-ocean ridges. PMID:25209311

  17. Lead-cooled fast reactor use in future equilibrium energy production

    SciTech Connect

    Sekimoto, Hiroshi; Kuznetsov, V.V.

    1994-12-31

    The design of a lead cooled fast reactor is discussed. In previous works, general characteristics of future nuclear equilibrium energy utilization have been investigated where the toxic radioactive materials are confined in a nuclear center. Natural uranium and/or thorium, are supplied to the center as a fuel fed to fission reactors. All of the actinides are recycled in the reactor. The end products of the heavy-isotope decay series (lead and bismuth) and the stable fission products are taken out of the center. The discharged short- and middle-life fission products are cooled until they die out in the center, then the previously described process is used.

  18. Ultra fast cooling of hot steel plate by air atomized spray with salt solution

    NASA Astrophysics Data System (ADS)

    Mohapatra, Soumya S.; Ravikumar, Satya V.; Jha, Jay M.; Singh, Akhilendra K.; Bhattacharya, Chandrima; Pal, Surjya K.; Chakraborty, Sudipto

    2014-05-01

    In the present study, the applicability of air atomized spray with the salt added water has been studied for ultra fast cooling (UFC) of a 6 mm thick AISI-304 hot steel plate. The investigation includes the effect of salt (NaCl and MgSO4) concentration and spray mass flux on the cooling rate. The initial temperature of the steel plate before the commencement of cooling is kept at 900 °C or above, which is usually observed as the "finish rolling temperature" in the hot strip mill of a steel plant. The heat transfer analysis shows that air atomized spray with the MgSO4 salt produces 1.5 times higher cooling rate than atomized spray with the pure water, whereas air atomized spray with NaCl produces only 1.2 times higher cooling rate. In transition boiling regime, the salt deposition occurs which causes enhancement in heat transfer rate by conduction. Moreover, surface tension is the governing parameter behind the vapour film instability and this length scale increases with increase in surface tension of coolant. Overall, the achieved cooling rates produced by both types of salt added air atomized spray are found to be in the UFC regime.

  19. Sensitivity Analysis of Reprocessing Cooling Times on Light Water Reactor and Sodium Fast Reactor Fuel Cycles

    SciTech Connect

    R. M. Ferrer; S. Bays; M. Pope

    2008-04-01

    The purpose of this study is to quantify the effects of variations of the Light Water Reactor (LWR) Spent Nuclear Fuel (SNF) and fast reactor reprocessing cooling time on a Sodium Fast Reactor (SFR) assuming a single-tier fuel cycle scenario. The results from this study show the effects of different cooling times on the SFR’s transuranic (TRU) conversion ratio (CR) and transuranic fuel enrichment. Also, the decay heat, gamma heat and neutron emission of the SFR’s fresh fuel charge were evaluated. A 1000 MWth commercial-scale SFR design was selected as the baseline in this study. Both metal and oxide CR=0.50 SFR designs are investigated.

  20. Measurements of thermal-hydraulic parameters in liquid-metal-cooled fast-breeder reactors

    SciTech Connect

    Sackett, J.I.

    1983-01-01

    This paper discusses instrumentation for liquid-metal-cooled fast breeder reactors (LMFBR's). Included is instrumentation to measure sodium flow, pressure, temperature, acoustic noise, sodium purity, and leakage. The paper identifies the overall instrumentation requirements for LMFBR's and those aspects of instrumentation which are unique or of special concern to LMFBR systems. It also gives an overview of the status of instrument design and performance.

  1. Studies on Materials for Heavy-Liquid-Metal-Cooled Reactors in Japan

    SciTech Connect

    Minoru Takahashi; Masayuki Igashira; Toru Obara; Hiroshi Sekimoto; Kenji Kikuchi; Kazumi Aoto; Teruaki Kitano

    2002-07-01

    Recent studies on materials for the development of lead-bismuth (Pb-Bi)-cooled fast reactors (FR) and accelerator-driven sub-critical systems (ADS) in Japan are reported. The measurement of the neutron cross section of Bi to produce {sup 210}Po, the removal experiment of Po contamination and steel corrosion test in Pb-Bi flow were performed in Tokyo Institute of Technology. A target material corrosion test was performed in the project of Transmutation Experimental Facility for ADS in Japan Atomic Energy Research Institute (JAERI). Steel corrosion test was started in Mitsui Engineering and Shipbuilding Co., LTD (MES). The feasibility study for FR cycle performed in Japan Nuclear Cycle Institute (JNC) are described. (authors)

  2. Development of fast cooling pulsed magnets at the Wuhan National High Magnetic Field Center

    SciTech Connect

    Peng, Tao; Sun, Quqin; Zhao, Jianlong; Jiang, Fan; Li, Liang; Xu, Qiang; Herlach, Fritz

    2013-12-15

    Pulsed magnets with fast cooling channels have been developed at the Wuhan National High Magnetic Field Center. Between the inner and outer sections of a coil wound with a continuous length of CuNb wire, G10 rods with cross section 4 mm × 5 mm were inserted as spacers around the entire circumference, parallel to the coil axis. The free space between adjacent rods is 6 mm. The liquid nitrogen flows freely in the channels between these rods, and in the direction perpendicular to the rods through grooves provided in the rods. For a typical 60 T pulsed magnetic field with pulse duration of 40 ms, the cooling time between subsequent pulses is reduced from 160 min to 35 min. Subsequently, the same technology was applied to a 50 T magnet with 300 ms pulse duration. The cooling time of this magnet was reduced from 480 min to 65 min.

  3. Thermal Hydraulic Challenges of Gas Cooled Fast Reactors with Passive Safety Features

    SciTech Connect

    Michael Pope; Jeong-Ik Lee; Pavel Hejzlar; Michael J. Driscoll

    2009-05-01

    Transient response of a Gas cooled Fast Reactor (GFR) coupled to a recompression supercritical CO2 (S-CO2) power conversion system (PCS) in a direct cycle to Loss of Coolant and Loss of Generator Load Accidents is analyzed using RELAP5-3D. A number of thermal hydraulic challenges for GFR design are pointed out as the designers strive to accommodate cooling of the high power density core of a fast reactor by a gas with its inherently low heat transfer capability, in particular under post LOCA events when system pressure is lost and when reliance on passive decay heat removal is emphasized. Although it is possible to design a S-CO2 cooled GFR that can survive LOCA by cooling the core through natural circulating loops between the core and elevated emergency cooling heat exchangers, it is not an attractive approach because of various bypass paths that can, depending on break location, degrade core cooling. Moreover, natural circulation gas loops can operate in deteriorated heat transfer regimes with substantial reduction of heat transfer coefficient: as low as 30% of forced convection values, and data and correlations in these regimes carry large uncertainties. Therefore, reliable battery powered blowers for post-LOCA decay heat removal (DHR) that provide flow in well defined regimes with low uncertainty, and can be easily over-designed to accommodate bypass flows were selected. The results confirm that a GFR with such a DHR system and negative coolant void worth can withstand LOCA with and without scram as well as loss of electrical load without exceeding core temperature and turbomachinery overspeed limits.

  4. Development of a neutronics calculation method for designing commercial type Japanese sodium-cooled fast reactor

    SciTech Connect

    Takeda, T.; Shimazu, Y.; Hibi, K.; Fujimura, K.

    2012-07-01

    Under the R and D project to improve the modeling accuracy for the design of fast breeder reactors the authors are developing a neutronics calculation method for designing a large commercial type sodium- cooled fast reactor. The calculation method is established by taking into account the special features of the reactor such as the use of annular fuel pellet, inner duct tube in large fuel assemblies, large core. The Verification and Validation, and Uncertainty Qualification (V and V and UQ) of the calculation method is being performed by using measured data from the prototype FBR Monju. The results of this project will be used in the design and analysis of the commercial type demonstration FBR, known as the Japanese Sodium fast Reactor (JSFR). (authors)

  5. Spectroscopic and microscopic investigation of the corrosion of D-9 stainless steel by lead bismuth eutectic (LBE) at elevated temperatures. Initiation of thick oxide formation

    NASA Astrophysics Data System (ADS)

    Johnson, Allen L.; Koury, Dan; Welch, Jenny; Ho, Thao; Sidle, Stacy; Harland, Chris; Hosterman, Brian; Younas, Umar; Ma, Longzhou; Farley, John W.

    2008-06-01

    Corrosion of 316/316L stainless steel by lead-bismuth eutectic (LBE) at elevated temperature was investigated by examination of samples after 1000, 2000, and 3000 h of exposure at 550 °C, using SEM, XPS with sputter depth profiling, and TEM. The process by which localized oxide failure becomes extensive thick oxide formation was investigated. Under our experimental conditions, iron was observed to migrate outward while chromium did not migrate above the original metal surface. The thin oxide layer on the D-9 sample resembled 316L cold-rolled samples, while the thick oxide on D-9 resembled annealed 316L oxide. With continued exposure, thick oxide grew to cover the entire surface.

  6. Electrochemical and Mechanical Behavior of Lead-Silver and Lead-Bismuth Casting Alloys for Lead-Acid Battery Components

    NASA Astrophysics Data System (ADS)

    Osório, Wislei R.; Peixoto, Leandro C.; Garcia, Amauri

    2015-09-01

    The present study focuses on the interrelation of microstructure, mechanical properties, and corrosion resistance of Pb-Ag and Pb-Bi casting alloys, which can be used in the manufacture of lead-acid battery components, as potential alternatives to alloys currently used. A water-cooled solidification system is used, in which vertical upward directional solidification is promoted permitting a wide range of microstructures to be investigated. Correlations between microstructural arrays, tensile strengths, and corrosion resistances of Pb-1 wt pct Ag, Pb-2.5 wt pct Ag, Pb-1 wt pct Bi, and Pb-2.5 wt pct Bi alloys are envisaged. It is shown that a compromise between corrosion resistance (represented by the corrosion current density) and mechanical properties (represented by the ultimate tensile strength) can be obtained. Comparisons between specific strengths and mechanical/corrosion ratios are also made. It is also shown that, for microstructures solidified under cooling rates higher than 10 K/s, the Pb-Ag alloys exhibit higher specific strength and mechanical/corrosion ratio. In contrast, for casting processes in which the cooling rates are lower than 5 K/s, the dilute Pb-Bi alloy ( i.e., 1 wt pct Bi) is shown to have more appropriate requirements for lead-acid battery components. Comparisons between specific strengths, mechanical/corrosion ratio, and relative weight and cost with Pb-Sn and Pb-Sb alloys are also made.

  7. SSTAR: The U.S. Lead-Cooled Fast Reactor (LFR)

    SciTech Connect

    Smith, C F; Halsey, W G; Brown, N W; Sienicki, J J; Moisseytsev, A; Wade, D C

    2007-09-25

    It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10-700 MWe) on small or immature grids in developing nations. Most recently, the Global Nuclear Energy Partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the Small Secure Transportable Autonomous Reactor (SSTAR) reactor has been under ongoing development under the U.S. Generation IV Nuclear Energy Systems Initiative. It a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation aims, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the U.S. Generation IV Lead-cooled Fast Reactor system.

  8. ATWS Transients for the 2400 MWt Gas-Cooled Fast Reactor

    SciTech Connect

    Cheng,L.Y.; Ludewig, H.

    2007-08-05

    Reactivity transients have been analyzed with an updated RELAPS-3D (ver. 2.4.2) system model of the pin core design for the 2400MWt gas-cooled fast reactor (GCFR). Additional reactivity parameters were incorporated in the RELAP5 point-kinetics model to account for reactivity feedbacks due to axial and radial expansion of the core, fuel temperature changes (Doppler effect), and pressure changes (helium density changes). Three reactivity transients without scram were analyzed and the incidents were initiated respectively by reactivity ramp, loss of load, and depressurization. During the course of the analysis the turbine bypass model for the power conversion unit (PCU) was revised to enable a better utilization of forced flow cooling after the PCU is tripped. The analysis of the reactivity transients demonstrates the significant impact of the PCU on system pressure and core flow. Results from the modified turbine bypass model suggest a success path for the GCFR to mitigate reactivity transients without scram.

  9. Generation IV nuclear energy system initiative. Pin core subassembly designfor the Gas-Cooled Fast Reactor.

    SciTech Connect

    Farmer, M. T.; Hoffman, E. A.; Pfeiffer, P. F.; Therios, I. U.

    2006-07-31

    The Gas-Cooled Fast Reactor (GFR) is one of six systems selected for viability assessment in the Generation IV program. It features a closed nuclear fuel cycle, consisting of a high-temperature helium-cooled fast spectrum reactor, coupled to a direct-cycle helium turbine for electricity production. The GFR combines the advances of fast spectrum systems with those of high-temperature systems. It was clear from the very beginning that GFR design should be driven by the objective to offer a complementary approach to liquid metal cooling. On this basis, CEA and the US DOE decided to collaborate on the pre-conceptual design of a GFR. This reactor design will provide a high level of safety and full recycling of the actinides, and will also be highly proliferation resistant and economically attractive. The status of this collaborative project is that two unit sizes, 600 MWt and 2400 MWt were selected as the focus of the design and safety studies. Researchers studied fuel forms, fuel assembly/element designs, core configurations, primary and balance-of-plant layouts, and safety approaches for both of these unit sizes. Results regarding the feasibility of this GFR design are encouraging. For example, sustainability and non-proliferation goals can be met and the proposed concept has attractive safety features. These features take advantage of the helium in terms of its neutronic quasi-transparency as well as the enhanced Doppler effect in connection with candidate fuel and structural materials. The current design trend is to consider high unit power for the GFR (2400 MWt), an attractive level for the power density (100 MW/m{sup 3}), and the implementation of an innovative plate type fuel or pin type sub-assembly with carbide-based actinide compounds and SiC-based structural materials. Work is still needed to refine the safety approach, to select the main system options, and to more definitively establish economic parameters.

  10. An Innovative Hybrid Loop-Pool Design for Sodium Cooled Fast Reactor

    SciTech Connect

    Haihua Zhao; Hongbin Zhang

    2007-11-01

    The existing sodium cooled fast reactors (SFR) have two types of designs – loop type and pool type. In the loop type design, such as JOYO (Japan) [1] and MONJU (Japan), the primary coolant is circulated through intermediate heat exchangers (IHX) external to the reactor tank. The major advantages of loop design include compactness and easy maintenance. The disadvantage is higher possibility of sodium leakage. In the pool type design such as EBR-II (USA), BN-600M(Russia), Superphénix (France) and European Fast Reactor [2], the reactor core, primary pumps, IHXs and direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) all are immersed in a pool of sodium coolant within the reactor vessel, making a loss of primary coolant extremely unlikely. However, the pool type design makes primary system large. In the latest ANL’s Advanced Burner Test Reactor (ABTR) design [3], the primary system is configured in a pool-type arrangement. The hot sodium at core outlet temperature in hot pool is separated from the cold sodium at core inlet temperature in cold pool by a single integrated structure called Redan. Redan provides the exchange of the hot sodium from hot pool to cold pool through IHXs. The IHXs were chosen as the traditional tube-shell design. This type of IHXs is large in size and hence large reactor vessel is needed.

  11. Inherent Prevention and Mitigation of Severe Accident Consequences in Sodium-Cooled Fast Reactors

    SciTech Connect

    Roald A. Wigeland; James E. Cahalan

    2011-04-01

    Safety challenges for sodium-cooled fast reactors include maintaining core temperatures within design limits and assuring the geometry and integrity of the reactor core. Due to the high power density in the reactor core, heat removal requirements encourage the use of high-heat-transfer coolants such as liquid sodium. The variation of power across the core requires ducted assemblies to control fuel and coolant temperatures, which are also used to constrain core geometry. In a fast reactor, the fuel is not in the most neutronically reactive configuration during normal operation. Accidents leading to fuel melting, fuel pin failure, and fuel relocation can result in positive reactivity, increasing power, and possibly resulting in severe accident consequences including recriticalities that could threaten reactor and containment integrity. Inherent safety concepts, including favorable reactivity feedback, natural circulation cooling, and design choices resulting in favorable dispersive characteristics for failed fuel, can be used to increase the level of safety to the point where it is highly unlikely, or perhaps even not credible, for such severe accident consequences to occur.

  12. Adaptive polynomial chaos techniques for uncertainty quantification of a gas cooled fast reactor transient

    SciTech Connect

    Perko, Z.; Gilli, L.; Lathouwers, D.; Kloosterman, J. L.

    2013-07-01

    Uncertainty quantification plays an increasingly important role in the nuclear community, especially with the rise of Best Estimate Plus Uncertainty methodologies. Sensitivity analysis, surrogate models, Monte Carlo sampling and several other techniques can be used to propagate input uncertainties. In recent years however polynomial chaos expansion has become a popular alternative providing high accuracy at affordable computational cost. This paper presents such polynomial chaos (PC) methods using adaptive sparse grids and adaptive basis set construction, together with an application to a Gas Cooled Fast Reactor transient. Comparison is made between a new sparse grid algorithm and the traditionally used technique proposed by Gerstner. An adaptive basis construction method is also introduced and is proved to be advantageous both from an accuracy and a computational point of view. As a demonstration the uncertainty quantification of a 50% loss of flow transient in the GFR2400 Gas Cooled Fast Reactor design was performed using the CATHARE code system. The results are compared to direct Monte Carlo sampling and show the superior convergence and high accuracy of the polynomial chaos expansion. Since PC techniques are easy to implement, they can offer an attractive alternative to traditional techniques for the uncertainty quantification of large scale problems. (authors)

  13. Mechanical performance and microstructure array of as-cast lead-silver and lead-bismuth alloys

    NASA Astrophysics Data System (ADS)

    Osório, Wislei R.; Bortolozo, Ausdinir D.; Peixoto, Leandro C.; Garcia, Amauri

    2014-12-01

    The aim of this study is to establish correlations between mechanical properties of Pb-Ag and Pb-Bi alloys and parametric features of their as-cast microstructures, as well as to develop a comparative analysis with the corresponding properties of Pb-Sn alloys considering applications of these alloys in the manufacture of Pb-acid battery components. A wide range of microstructures are obtained using an upward water-cooled directional solidification system. Ultimate (UTS) and yield tensile strengths (YS) and elongation are experimentally determined as a function of cellular and dendritic spacings, and Hall-Petch type experimental equations are proposed relating UTS to these microstructure parameters. Despite the higher specific strengths of Pb-Ag alloys, as compared with those of Pb-Bi and Pb-Sn alloys, their corresponding relative costs are the highest of all Pb-based alloys examined. It is found that the Pb-Bi and Pb-Sn alloys examined have similar specific strengths and relative costs.

  14. LSP simulations of fast ions slowing down in cool magnetized plasma

    NASA Astrophysics Data System (ADS)

    Evans, Eugene S.; Cohen, Samuel A.

    2015-11-01

    In MFE devices, rapid transport of fusion products, e.g., tritons and alpha particles, from the plasma core into the scrape-off layer (SOL) could perform the dual roles of energy and ash removal. Through these two processes in the SOL, the fast particle slowing-down time will have a major effect on the energy balance of a fusion reactor and its neutron emissions, topics of great importance. In small field-reversed configuration (FRC) devices, the first-orbit trajectories of most fusion products will traverse the SOL, potentially allowing those particles to deposit their energy in the SOL and eventually be exhausted along the open field lines. However, the dynamics of the fast-ion energy loss processes under conditions expected in the FRC SOL, where the Debye length is greater than the electron gyroradius, are not fully understood. What modifications to the classical slowing down rate are necessary? Will instabilities accelerate the energy loss? We use LSP, a 3D PIC code, to examine the effects of SOL plasma parameters (density, temperature and background magnetic field strength) on the slowing down time of fast ions in a cool plasma with parameters similar to those expected in the SOL of small FRC reactors. This work supported by DOE contract DE-AC02-09CH11466.

  15. A 100 MWe advanced sodium-cooled fast reactor core concept

    SciTech Connect

    Kim, T. K.; Grandy, C.; Hill, R. N.

    2012-07-01

    An Advanced sodium-cooled Fast Reactor core concept (AFR-100) was developed targeting a small electrical grid to be transportable to the plant site and operable for a long time without frequent refueling. The reactor power rating was strategically decided to be 100 MWe, and the core barrel diameter was limited to 3.0 m for transportability. The design parameters were determined by relaxing the peak fast fluence limit and bulk coolant outlet temperature to beyond irradiation experience assuming that advanced cladding and structural materials developed under US-DOE programs would be available when the AFR-100 is deployed. With a de-rated power density and U-Zr binary metallic fuel, the AFR-100 can maintain criticality for 30 years without refueling. The average discharge burnup of 101 MWd/kg is comparable to conventional design values, but the peak discharge fast fluence of {approx}6x10{sup 23} neutrons/cm{sup 2} is beyond the current irradiation experiences with HT-9 cladding. The evaluated reactivity coefficients provide sufficient negative feedbacks and the reactivity control systems provide sufficient shutdown margins. The integral reactivity parameters obtained from quasi-static reactivity balance analysis indicate that the AFR-100 meets the sufficient conditions for acceptable asymptotic core outlet temperature following postulated unprotected accidents. Additionally, the AFR-100 has sufficient thermal margins by grouping the fuel assemblies into eight orifice zones. (authors)

  16. Creep, creep-rupture tests of Al-surface-alloyed T91 steel in liquid lead bismuth at 500 and 550 °C

    NASA Astrophysics Data System (ADS)

    Weisenburger, A.; Jianu, A.; An, W.; Fetzer, R.; Del Giacco, Mattia; Heinzel, A.; Müller, G.; Markov, V. G.; Kasthanov, A. D.

    2012-12-01

    Surface layers made of FeCrAl alloys on T91 steel have shown their capability as corrosion protection barriers in lead bismuth. Pulsed electron beam treatment improves the density and more over the adherence of such layers. After the treatment of previously deposited coatings a surface graded material is achieved with a metallic bonded interface. Creep-rupture tests of T91 in lead-alloy at 550 °C reveal significant reduced creep strength of non-modified T91 test specimens. Oxide scales protecting the steels from attacks of the liquid metal will crack at a certain strain leading to a direct contact between the steel and the liquid metal. The negative influence of the lead-alloy on the creep behavior of non-modified T91 is stress dependent, but below a threshold stress value of 120 MPa at 550 °C this influence becomes almost negligible. At 500 °C and stress values of 200 MPa and 220 MPa the creep rates are comparable between them and significantly lower than creep rates at 180 MPa of original T91 in air at 550 °C. No signs of LBE influence are detected. The surface modified specimens tested at high stress levels instead had creep-rupture times similar to T91 (original state) tested in air. The thin oxide layers formed on the surface modified steel samples are less susceptible to crack formation and therefore to lead-alloy enhanced creep.

  17. Influence of liquid lead and lead-bismuth eutectic on tensile, fatigue and creep properties of ferritic/martensitic and austenitic steels for transmutation systems

    NASA Astrophysics Data System (ADS)

    Gorse, D.; Auger, T.; Vogt, J.-B.; Serre, I.; Weisenburger, A.; Gessi, A.; Agostini, P.; Fazio, C.; Hojna, A.; Di Gabriele, F.; Van Den Bosch, J.; Coen, G.; Almazouzi, A.; Serrano, M.

    2011-08-01

    In this paper, the tensile, fatigue and creep properties of the Ferritic/Martensitic (F/M) steel T91 and of the Austenitic Stainless (AS) Steel 316L in lead-bismuth eutectic (LBE) or lead, obtained in the different organizations participating to the EUROTRANS-DEMETRA project are reviewed. The results show a remarkable consistency, referring to the variety of metallurgical and surface state conditions studied. Liquid Metal Embrittlement (LME) effects are shown, remarkable on heat-treated hardened T91 and also on corroded T91 after long-term exposure to low oxygen containing Liquid Metal (LM), but hardly visible on passive or oxidized smooth T91 specimens. For T91, the ductility trough was estimated, starting just above the melting point of the embrittler ( TM,E = 123.5 °C for LBE, 327 °C for lead) with the ductility recovery found at 425 °C. LME effects are weaker on 316L AS steel. Liquid Metal Assisted Creep (LMAC) effects are reported for the T91/LBE system at 550 °C, and for the T91/lead system at 525 °C. Today, if the study of the LME effects on T91 and 316L in LBE or lead can be considered well documented, in contrast, complementary investigations are necessary in order to quantify the LMAC effects in these systems, and determine rigorously the threshold creep conditions.

  18. Assessment of the influence of surface finishing and weld joints on the corrosion/oxidation behaviour of stainless steels in lead bismuth eutectic

    NASA Astrophysics Data System (ADS)

    Martín-Muñoz, F. J.; Soler-Crespo, L.; Gómez-Briceño, D.

    2011-09-01

    The objective of this paper is to gain some insight into the influence of the surface finishing in the oxidation/corrosion behaviour of 316L and T91 steels in lead bismuth eutectic (LBE). Specimens of both materials with different surface states were prepared (as-received, grinded, grinded and polished, and electrolitically polished) and oxidation tests were carried out at 775 and 825 K from 100 to 2000 h for two different oxygen concentrations and for H 2/H 2O molar ratios of 3 and 0.03. The general conclusion for these tests is that the effect of surface finishing on the corrosion/protection processes is not significant under the tested conditions. In addition the behaviour of weld joints, T91-T91 Tungsten Inert Gas (TIG) and T91-316L have been also studied under similar conditions. The conclusions are that, whereas T91-T91 welded joint shows the same corrosion properties as the parent materials for the conditions tested, AISI 316L-T91 welded joint, present an important dissolution over seam area that it associated to the electrode 309S used for the fabrication process.

  19. Effect of irradiation defects on the corrosion behaviors of steels exposed to lead bismuth eutectic in ADS: a first-principles study.

    PubMed

    Zhang, Yange; You, Yu-Wei; Li, Dong-Dong; Xu, Yichun; Liu, C S; Pan, B C; Wang, Zhiguang

    2015-05-14

    In accelerator driven systems (ADSs), steels will suffer not only from the irradiation damage produced by protons or neutrons, but also from the dissolution corrosion induced by the liquid lead-bismuth eutectic (LBE). In this work we investigate the interactions between LBE atoms (Pb, Bi) and the irradiation induced defects X (X is helium, vacancy or divacancy) in α-Fe based on first-principles calculations. It is found that LBE atoms repulse each other without irradiation defects, while they aggregate easily with the defects to form X-Pbn and X-Bin complexes. This indicates that the irradiation defects could promote the aggregation of LBE atoms in iron, especially Bi atoms. The total binding energies of the X-Pbn and X-Bin complexes increase with the number of Pb and Bi atoms, respectively. The origin of the total binding energies of the complexes is further discussed via the electronic structures and the distortion of the crystalline lattice. Finally, the concentration evolutions of the Vac-(Bi)n complexes and unbound vacancies with temperature are predicted by the mass action analysis. This work provides important information for the synergistic effect of irradiation and LBE corrosion on the steels in the ADSs, which can be used as basic parameters for further study. PMID:25891773

  20. Silicon-containing ferritic/martensitic steel after exposure to oxygen-containing flowing lead-bismuth eutectic at 450 and 550 °C

    NASA Astrophysics Data System (ADS)

    Schroer, Carsten; Koch, Verena; Wedemeyer, Olaf; Skrypnik, Aleksandr; Konys, Jürgen

    2016-02-01

    A ferritic/martensitic (f/m) steel with 9 and 3 mass% of chromium (Cr) and silicon (Si), respectively, was tested on performance in flowing lead-bismuth eutectic (LBE) at 450 and 550 °C, each at concentrations of solved oxygen of both 10-7 and 10-6 mass%. The 9Cr-3Si steel generally exhibits the same basic corrosion modes as other f/m materials with 9 mass% Cr and typically lower Si content, namely Steel T91. The Si-rich steel shows an overall improved performance in comparison to T91 at 450 °C and 10-7 mass% solved oxygen, but especially at 450 °C and 10-6 mass% solved oxygen. The advantage of higher Si-content in 9Cr steel is less clear at 550 °C. Especially high oxygen content in flowing LBE at 550 °C, between >10-6 mass% and oxygen saturation, seems detrimental for the high-Si material in respect of the initiation and progress of a solution-based corrosion.

  1. Analysis of bi-layer oxide on austenitic stainless steel, 316L, exposed to Lead-Bismuth Eutectic (LBE) by X-ray Photoelectron Spectroscopy (XPS)

    NASA Astrophysics Data System (ADS)

    Koury, D.; Johnson, A. L.; Ho, T.; Farley, J. W.

    2013-09-01

    Corrosion of the austenitic stainless steel alloy 316L by Lead-Bismuth Eutectic (LBE) was studied using X-ray Photoelectron Spectroscopy (XPS) with Sputter-Depth Profiling (SDP), and compared to data taken by Scanning Electron Microscopy (SEM) and Energy Dispersive X-rays (EDXs). Exposed and unexposed samples were compared. Annealed 316L samples, exposed to LBE for durations of 1000, 2000 and 3000 h, developed bi-layer oxides up to 30 μm thick. Analysis of the charge-states of the 2p3/2 peaks of iron, chromium, and nickel in the oxide layers reveal an inner layer consisting of iron and chromium oxides (likely spinel-structured) and an outer layer consisting of iron oxides (Fe3O4). Cold-rolled 316L samples, exposed for the same durations, form a chromium-rich, thin (⩽1 μm) oxide with some oxidized iron in the outermost ˜200 nm of the oxide layer. This is the first experiment to investigate what components of the 316L are oxidized by LBE exposure. It is shown here that nickel is metallic in the inner layer.

  2. Spectroscopic and microscopic investigation of the corrosion of 316/316L stainless steel by lead-bismuth eutectic (LBE) at elevated temperatures: importance of surface preparation

    NASA Astrophysics Data System (ADS)

    Johnson, Allen L.; Parsons, Denise; Manzerova, Julia; Perry, Dale L.; Koury, Dan; Hosterman, Brian; Farley, John W.

    2004-07-01

    The corrosion of steel by lead-bismuth eutectic (LBE) is an important issue in proposed nuclear transmutation schemes. Russian scientists at the IPPE exposed steel samples to oxygen-controlled LBE at temperatures up to 823 K and exposure times up to 3000 h. We have characterized these post-exposure steel samples and unexposed controls, using scanning electron microscopy (SEM), energy-dispersive X-ray analysis (EDAX) and X-ray photoelectron spectroscopy (XPS). Previous researchers have investigated the corrosion by LBE of steel of varying composition. In the present work, we compared two samples having the same composition (standard nuclear grade 316/316L) but different surface preparation: a cold-rolled sample was compared with an annealed sample. The cold-rolled sample had an order of magnitude less corrosion (i.e., both lower oxidation and less weight change) than the annealed sample. Sputter depth profiling of the exposed annealed sample and cold-rolled sample showed a marked difference in oxide layer composition between the annealed and cold-rolled samples. The annealed sample showed a complex oxide structure (iron oxide over chromium/iron oxide mixtures) of tens of microns thickness, while the cold-rolled sample was covered with a rather simple, primarily chromium oxide layer of ˜1 μm thickness.

  3. 2400MWt GAS-COOLED FAST REACTOR DHR STUDIES STATUS UPDATE.

    SciTech Connect

    CHENG,L.Y.; LUDEWIG, H.

    2007-06-01

    A topical report on demonstrating the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400MWt GEN-IV gas-cooled fast reactor was published in March 2006. The analysis was performed with the system code RELAP5-3D (version 2.4.1.1a) and the model included the full complement of the power conversion unit (PCU): heat exchange components (recuperator, precooler, intercooler) and rotating machines (turbine, compressor). A re-analysis of the success case in Ref is presented in this report. The case was redone to correct unexpected changes in core heat structure temperatures when the PCU model was first integrated with the reactor model as documented in Ref [1]. Additional information on the modeling of the power conversion unit and the layout of the heat exchange components is provided in Appendix A.

  4. Formation of snowflake domains during fast cooling of lithium tantalate crystals

    NASA Astrophysics Data System (ADS)

    Shur, V. Ya.; Kosobokov, M. S.; Mingaliev, E. A.; Kuznetsov, D. K.; Zelenovskiy, P. S.

    2016-04-01

    Formation of the original dendrite snowflake-shape domains during fast cooling after heating above phase transition temperature by pulse laser irradiation was revealed in congruent lithium tantalate crystals. The effect was attributed to polarization reversal under the action of spatially nonuniform pyroelectric field. Two stages of the domain shape evolution at the surface were separated: (1) growth of circular domains by sideways motion of the domain walls and (2) backswitching leading to formation of the snowflake domains. The simulated spatial distribution of the pyroelectric field in regular two-dimensional structure was used for an explanation of the obtained results. The backswitching process in the surface layer has been attributed to change of the sign of the pyroelectric field at the domain wall. The snowflake domain shape is caused by the formation of isolated nanodomain fingers and hampering of their merging.

  5. Method of detecting leakage of reactor core components of liquid metal cooled fast reactors

    DOEpatents

    Holt, Fred E.; Cash, Robert J.; Schenter, Robert E.

    1977-01-01

    A method of detecting the failure of a sealed non-fueled core component of a liquid-metal cooled fast reactor having an inert cover gas. A gas mixture is incorporated in the component which includes Xenon-124; under neutron irradiation, Xenon-124 is converted to radioactive Xenon-125. The cover gas is scanned by a radiation detector. The occurrence of 188 Kev gamma radiation and/or other identifying gamma radiation-energy level indicates the presence of Xenon-125 and therefore leakage of a component. Similarly, Xe-126, which transmutes to Xe-127 and Kr-84, which produces Kr-85.sup.m can be used for detection of leakage. Different components are charged with mixtures including different ratios of isotopes other than Xenon-124. On detection of the identifying radiation, the cover gas is subjected to mass spectroscopic analysis to locate the leaking component.

  6. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    SciTech Connect

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.

  7. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    SciTech Connect

    Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

    2014-09-30

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  8. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

    2014-09-01

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  9. Impact of nuclear data on sodium-cooled fast reactor calculations

    NASA Astrophysics Data System (ADS)

    Aures, Alexander; Bostelmann, Friederike; Zwermann, Winfried; Velkov, Kiril

    2016-03-01

    Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors.

  10. Safety design approach for external events in Japan sodium-cooled fast reactor

    SciTech Connect

    Yamano, H.; Kubo, S.; Tani, A.; Nishino, H.; Sakai, T.

    2012-07-01

    This paper describes a safety design approach for external events in the design study of Japan sodium-cooled fast reactor. An emphasis is introduction of a design extension external condition (DEEC). In addition to seismic design, other external events such as tsunami, strong wind, abnormal temperature, etc. were addressed in this study. From a wide variety of external events consisting of natural hazards and human-induced ones, a screening method was developed in terms of siting, consequence, frequency to select representative events. Design approaches for these events were categorized on the probabilistic, statistical and deterministic basis. External hazard conditions were considered mainly for DEECs. In the probabilistic approach, the DEECs of earthquake, tsunami and strong wind were defined as 1/10 of exceedance probability of the external design bases. The other representative DEECs were also defined based on statistical or deterministic approaches. (authors)

  11. A Subcritical, Gas-Cooled Fast Transmutation Reactor with a Fusion Neutron Source

    SciTech Connect

    Stacey, W.M.; Beavers, V.L.; Casino, W.A.; Cheatham, J.R.; Friis, Z.W.; Green, R.D.; Hamilton, W.R.; Haufler, K.W.; Hutchinson, J.D.; Lackey, W.J.; Lorio, R.A.; Maddox, J.W.; Mandrekas, J.; Manzoor, A.A.; Noelke, C.A.; Oliveira, C. de; Park, M.; Tedder, D.W.; Terry, M.R.; Hoffman, E.A.

    2005-05-15

    A design is presented for a subcritical, He-cooled fast reactor, driven by a tokamak D-T fusion neutron source, for the transmutation of spent nuclear fuel (SNF). The reactor is fueled with coated transuranic (TRU) particles and is intended for the deep-burn (>90%) transmutation of the TRUs in SNF without reprocessing of the coated fuel particles. The reactor design is based on the materials, fuel, and separations technologies under near-term development in the U.S. Department of Energy (DOE) Nuclear Energy Program and on the plasma physics and fusion technologies under near-term development in the DOE Fusion Energy Sciences Program, with the objective of intermediate-term ({approx}2040) deployment. The physical and performance characteristics and research and development requirements of such a reactor are described.

  12. CFD Modeling of Sodium-Oxide Deposition in Sodium-Cooled Fast Reactor Compact Heat Exchangers

    SciTech Connect

    Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

    2015-09-02

    The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HX channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.

  13. Interim Status Report on the Design of the Gas-Cooled Fast Reactor (GFR)

    SciTech Connect

    Weaver, K. D.

    2005-01-31

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report outlines the current design status of the GFR, and includes work done in the areas mentioned above.

  14. Interim Status Report on the Design of the Gas-Cooled Fast Reactor (GFR)

    SciTech Connect

    Kevan D. Weaver

    2005-01-01

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850ºC at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report outlines the current design status of the GFR, and includes work done in the areas mentioned above.

  15. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    DOE PAGESBeta

    Cheng, Lap-Yan; Wei, Thomas Y. C.

    2009-01-01

    The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow weremore » evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.« less

  16. Dissolution and oxidation behaviour of various austenitic steels and Ni rich alloys in lead-bismuth eutectic at 520 °C

    NASA Astrophysics Data System (ADS)

    Roy, Marion; Martinelli, Laure; Ginestar, Kevin; Favergeon, Jérôme; Moulin, Gérard

    2016-01-01

    Ten austenitic steels and Ni rich alloys were tested in static lead-bismuth eutectic (LBE) at 520 °C in order to obtain a selection of austenitic steels having promising corrosion behaviour in LBE. A test of 1850 h was carried out with a dissolved oxygen concentration between 10-9 and 5 10-4 g kg-1. The combination of thermodynamic of the studied system and literature results leads to the determination of an expression of the dissolved oxygen content in LBE as a function of temperature: RT(K)ln[O](wt%) = -57584/T(K) -55.876T(K) + 254546 (R is the gas constant in J mol-1 K-1). This relation can be considered as a threshold of oxygen content above which only oxidation is observed on the AISI 316L and AISI 304L austenitic alloys in static LBE between 400 °C and 600 °C. The oxygen content during the test leads to both dissolution and oxidation of the samples during the first 190 h and leads to pure oxidation for the rest of the test. Results of mixed oxidation and dissolution test showed that only four types of corrosion behaviour were observed: usual austenitic steels and Ni rich alloys behaviour including the reference alloy 17Cr-12Ni-2.5Mo (AISI 316LN), the 20Cr-31Ni alloy one, the Si containing alloy one and the Al containing alloy one. According to the proposed criteria of oxidation and dissolution kinetics, silicon rich alloys and aluminum rich alloy presented a promising corrosion behaviour.

  17. A physics study for negative void reactivity in compact supercritical CO{sub 2}-cooled fast reactor

    SciTech Connect

    Kim, Y.; Hartanto, D.; Lee, J. I.

    2013-07-01

    A compact S-CO{sub 2}-cooled fast reactor which has negative Coolant Void Reactivity (CVR) has been investigated. A negative CVR is important for the gas cooled fast reactor as an inherent safety mechanism to prevent the sudden positive reactivity insertion when the loss of coolant accident happens. An alternative solution to reduce the CVR is investigated in this study by using O-17 instead of O-16 in UO{sub 2} fuel. By using O-17 in the fuel, it is found that the CVR can even be negative. Impacts of the radial reflector on the CVR are also evaluated for the small SCO{sub 2} cooled fast reactor in this study. We have considered a pure lead (Pb) reflector and a lead magnesium eutectic (LME) reflector as alternative radial reflectors of the S-CO 2-cooled fast reactor. It has been shown that, with the LME radial reflector, the CVR can be negative, while the pure lead reflector provides a slightly positive CVR. (authors)

  18. Cleaning residual NaK in the fast flux test facility fuel storage cooling system

    SciTech Connect

    Burke, T.M.; Church, W.R.; Hodgson, K.M.

    2008-01-15

    The Fast Flux Test Facility (FFTF), located on the U.S. Department of Energy's Hanford Reservation, is a liquid metal-cooled test reactor. The FFTF was constructed to support the U.S. Liquid Metal Fast Breeder Reactor Program. The bulk of the alkali metal (sodium and NaK) has been drained and will be stored onsite prior to final disposition. Residual NaK needed to be removed from the pipes, pumps, heat exchangers, tanks, and vessels in the Fuel Storage Facility (FSF) cooling system. The cooling system was drained in 2004 leaving residual NaK in the pipes and equipment. The estimated residual NaK volume was 76 liters in the storage tank, 1.9 liters in the expansion tank, and 19-39 liters in the heat transfer loop. The residual NaK volume in the remainder of the system was expected to be very small, consisting of films, droplets, and very small pools. The NaK in the FSF Cooling System was not radiologically contaminated. The portions of the cooling system to be cleaned were divided into four groups: 1. The storage tank, filter, pump, and associated piping; 2. The heat exchanger, expansion tank, and associated piping; 3. Argon supply piping; 4. In-vessel heat transfer loop. The cleaning was contracted to Creative Engineers, Inc. (CEI) and they used their superheated steam process to clean the cooling system. It has been concluded that during the modification activities (prior to CEI coming onsite) to prepare the NaK Cooling System for cleaning, tank T-914 was pressurized relative to the In-Vessel NaK Cooler and NaK was pushed from the tank back into the Cooler and that on November 6, 2005, when the gas purge through the In-Vessel NaK Cooler was increased from 141.6 slm to 283.2 slm, NaK was forced from the In-Vessel NaK Cooler and it contacted water in the vent line and/or scrubber. The gases from the reaction then traveled back through the vent line coating the internal surface of the vent line with NaK and NaK reaction products. The hot gases also exited the

  19. Analysis of the conceptual shielding design for the upflow Gas-Cooled Fast Breeder Reactor

    SciTech Connect

    Slater, C.O.; Reed, D.A.; Cramer, S.N.; Emmett, M.B.; Tomlinson, E.T.

    1981-01-01

    Conceptual Shielding Configuration III for the Gas-Cooled Fast Breeder Reactor (GCFR) was analyzed by performing global calculations of neutron and gamma-ray fluences and correcting the results as appropriate with bias factors from localized calculations. Included among the localized calculations were the radial and axial cell streaming calculations, plus extensive preliminary calculations and three final confirmation calculations of the plenum flow-through shields. The global calculations were performed on the GCFR mid-level and the lower and upper plenum regions. Calculated activities were examined with respect to the design constraint, if any, imposed on the particular activity. The spatial distributions of several activities of interest were examined with the aid of isoplots (i.e., symbols are used to describe a surface on which the activity level is everywhere the same). In general the results showed that most activities were below the respective design constraints. Only the total neutron fluence in the core barrel appeared to be marginal with the present reactor design. Since similar results were obtained for an earlier design, it has been proposed that the core barrel be cooled with inlet plenum gas to maintain it at a temperature low enough that it can withstand a higher fluence limit. Radiation levels in the prestressed concrete reactor vessel (PCRV) and liner appeared to be sufficiently below the design constraint that expected results from the Radial Shield Heterogeneity Experiment should not force any levels above the design constraint. A list was also made of a number of issues which should be examined before completion of the final shielding design.

  20. Prediction of engine performance and wall erosion due to film cooling for the 'fast track' ablative thrust chamber

    NASA Technical Reports Server (NTRS)

    Trinh, Huu P.

    1994-01-01

    Efforts have been made at the Propulsion Laboratory (MSFC) to design and develop new liquid rocket engines for small-class launch vehicles. Emphasis of the efforts is to reduce the engine development time with the use of conventional designs while meeting engine reliability criteria. Consequently, the engine cost should be reduced. A demonstrative ablative thrust chamber, called 'fast-track', has been built. To support the design of the 'fast-track' thrust chamber, predictions of the wall temperature and ablation erosion rate of the 'fast-track' thrust chamber have been performed using the computational fluid dynamics program REFLEQS (Reactive Flow Equation Solver). The analysis is intended to assess the amount of fuel to be used for film cooling so that the erosion rate of the chamber ablation does not exceed its allowable limit. In addition, the thrust chamber performance loss due to an increase of the film cooling is examined.

  1. Passive shutdown device for gas cooled fast reactor: Lithium injection module

    SciTech Connect

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

    2006-07-01

    In this paper a passive reactivity control system for a Gas Cooled Fast Reactor is proposed. The Generation IV GCFR features a core with a relatively high power density, and control of transients and adequate shutdown under accidental situations must be assured for safe operation. Using a passive shutdown device rules out the possibility of unprotected transients. The proposed devices work by the passive introduction of {sup 6}Li into the core (Lithium Injection Module). Control is by the outlet temperature of the coolant gas in the fuel assemblies, employing a freeze seal. The proposed devices can be integrated into the regular control assemblies. A total of four LIMs is proposed in the core. Thermohydraulic calculations were done using the CATHARE code for a 600 MWth GCFR, for 2 types of transients: a loss of flow, and a control rod withdrawal. The calculations show that activation of one LIM is sufficient to keep the reactor power bounded, while activation of all LIMs in the core will shut down the reactor. The passive LIM devices are able to exclude unprotected transients in the GCFR core. (authors)

  2. Initial Requirements for Gas-Cooled Fast Reactor (GFR) System Design, Performance, and Safety Analysis Models

    SciTech Connect

    Kevan D. Weaver; Thomas Y. C. Wei

    2004-08-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  3. Effects of Ultra Fast Cooling on Microstructure and Mechanical Properties of Pipeline Steels

    NASA Astrophysics Data System (ADS)

    Tian, Yong; Li, Qun; Wang, Zhao-dong; Wang, Guo-dong

    2015-09-01

    X70 (steel A) and X80 (steel B) pipeline steels were fabricated by ultra fast cooling (UFC). UFC processing improves not only ultimate tensile strength (UTS), yield strength (YS), yield ratio (YS/UTS), and total elongation of both steels, but also their Charpy absorbed energy ( A K) as well. The microstructures of both steels were all composed of quasi polygonal, acicular ferrite (AF), and granular bainite. MA islands (the mixtures of brittle martensite and residual austenite) are more finely dispersed in steel B, and the amount of AF in steel B is much more than that in steel A. The strength of steel B is higher than that of steel A. This is mainly attributed to the effect of the ferrite grain refinement which is resulted from UFC processing. The finely dispersed MA islands not only provide dispersion strengthening, but also reduce loss of impact properties to pipeline steels. UFC produces low-temperature transformation microstructures containing larger amounts of AFs. The presence of AF is a crucial factor in achieving desired mechanical properties for both steels. It is suggested that the toughness of the experimental steel increases with increasing the amount of AF.

  4. Multiple lead seal assembly for a liquid-metal-cooled fast-breeder nuclear reactor

    DOEpatents

    Hutter, Ernest; Pardini, John A.

    1977-03-15

    A reusable multiple lead seal assembly provides leak-free passage of stainless-steel-clad instrument leads through the cover on the primary tank of a liquid-metal-cooled fast-breeder nuclear reactor. The seal isolates radioactive argon cover gas and sodium vapor within the primary tank from the exterior atmosphere and permits reuse of the assembly and the stainless-steel-clad instrument leads. Leads are placed in flutes in a seal body, and a seal shell is then placed around the seal body. Circumferential channels in the body and inner surface of the shell are contiguous and together form a conduit which intersects each of the flutes, placing them in communication with a port through the wall of the seal shell. Liquid silicone rubber sealant is injected into the flutes through the port and conduit; the sealant fills the space in the flutes not occupied by the leads themselves and dries to a rubbery hardness. A nut, threaded onto a portion of the seal body not covered by the seal shell, jacks the body out of the shell and shears the sealant without damage to the body, shell, or leads. The leads may then be removed from the body. The sheared sealant is cleaned from the body, leads, and shell and the assembly may then be reused with the same or different leads.

  5. Conceptual design features of the Kalimer-600 sodium cooled fast reactor

    SciTech Connect

    Hahn, Dohee; Kim, Yeong-Il; Kim, Seong-O; Lee, Jae-Han; Lee, Yong-Bum; Jeong, Hae-Yong

    2007-07-01

    An advanced sodium cooled fast reactor concept, KALIMER-600, has been developed by the Korea Atomic Energy Research Institute to satisfy the Gen-IV technology goals of sustainability, safety and reliability, economics and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on a proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design were verified through a safety analysis of its bounding events. The results for various unprotected events imply that the KALIMER-600 design can accommodate all the analyzed ATWS events. This self-regulation capability of the power without a scram is mainly attributed to the inherent reactivity feedback mechanisms implemented in the metal fuel core design and completely passive decay heat removal system. (authors)

  6. ASTRID sodium cooled fast reactor: Program for improving in service inspection and repair

    SciTech Connect

    Jadot, F.; De Dinechin, G.; Augem, J. M.; Sibilo, J.

    2011-07-01

    In the frame of the CEA, EDF, AREVA coordinated research program for the development of Generation IV sodium-cooled fast reactors (SFR), the ASTRID project was launched in 2010. For the future prototype, the improvement of in-service inspection and repair (ISI and R) capabilities was identified as a major issue. Following the pluri-annual SFR research program, the ISI and R main R and D axes remain: i) improvement of the primary system conceptual design, ii) development of measurement and inspection techniques (continuous monitoring instrumentation and periodic inspection tools), iii) accessibility and associated robotics, and iv) development and validation of repair processes. Associated ISI and R needs are being defined through an iterative method between designers and instrumentation specialists: adaptation of the Design to ISI and R requirements, fission chamber development, validation of the ultrasonic and chemical transducers, of ultrasonic non destructive simulation, of acoustic surveillance, of laser repair intervention processes, of connected robotic equipment. Moreover, CEA, as leader of the ASTRID Project, is willing to find new contributors, partners or suppliers, in order to get innovative, diversified, exhaustive and efficient solutions. (authors)

  7. Definition of a Robust Supervisory Control Scheme for Sodium-Cooled Fast Reactors

    SciTech Connect

    Ponciroli, Roberto; Passerini, Stefano; Vilim, Richard B.

    2016-01-01

    In this work, an innovative control approach for metal-fueled Sodium-cooled Fast Reactors is proposed. With respect to the classical approach adopted for base-load Nuclear Power Plants, an alternative control strategy for operating the reactor at different power levels by respecting the system physical constraints is presented. In order to achieve a higher operational flexibility along with ensuring that the implemented control loops do not influence the system inherent passive safety features, a dedicated supervisory control scheme for the dynamic definition of the corresponding set-points to be supplied to the PID controllers is designed. In particular, the traditional approach based on the adoption of tabulated lookup tables for the set-point definition is found not to be robust enough when failures of the implemented SISO (Single Input Single Output) actuators occur. Therefore, a feedback algorithm based on the Reference Governor approach, which allows for the optimization of reference signals according to the system operating conditions, is proposed.

  8. Design Study of Small Lead-Cooled Fast Reactors Using SiC Cladding and Structure

    SciTech Connect

    Abu Khalid Rivai; Minoru Takahashi

    2006-07-01

    Effects of SiC cladding and structure on neutronics of reactor core for small lead-cooled fast reactors have been investigated analytically. The fuel of this reactor was uranium nitride with {sup 235}U enrichment of 11% in inner core and 13% in outer core. The reactors were designed by optimizing the use of natural uranium blanket and nitride fuel to prolong the fuel cycle. The fuels can be used without re-shuffling for 15 years. The coolant of this reactor was lead. A calculation was also conducted for steel cladding and structure type as comparison with SiC cladding and structure type. The results of calculation indicated that the neutron energy spectrum of the core using SiC was slightly softer than that using steel. The SiC type reactor was designed to have criticality at the beginning of cycle (BOC), although the steel type reactor could not have critical condition with the same size and geometry. In other words, the SiC type core can be designed smaller than the steel type core. The result of the design analysis showed that neutron flux distributions and power distribution was made flatter because the outer core enrichment was higher than inner core. The peak power densities could remain constant over the reactor operation. The consumption capability of uranium was quite high, i.e. 13% for 125 MWt reactor and 25% for 375 MWt reactor at EOC. (authors)

  9. Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors

    SciTech Connect

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Asiah, Nur; Shafii, M. Ali; Khairurrijal

    2010-12-23

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (k{sub eff}) is in almost linear relations with the change of the fuel volume to coolant ratio.

  10. Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors

    NASA Astrophysics Data System (ADS)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Asiah, Nur; Shafii, M. Ali

    2010-12-01

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (keff) is in almost linear relations with the change of the fuel volume to coolant ratio.

  11. Comparative analysis of thorium and uranium fuel for transuranic recycle in a sodium cooled Fast Reactor

    SciTech Connect

    C. Fiorina; N. E. Stauff; F. Franceschini; M. T. Wenner; A. Stanculescu; T. K. Kim; A. Cammi; M. E. Ricotti; R. N. Hill; T. A. Taiwo; M. Salvatores

    2013-12-01

    The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associated with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232.

  12. Fast optical cooling of nanomechanical cantilever with the dynamical Zeeman effect.

    PubMed

    Zhang, Jian-Qi; Zhang, Shuo; Zou, Jin-Hua; Chen, Liang; Yang, Wen; Li, Yong; Feng, Mang

    2013-12-01

    We propose an efficient optical electromagnetically induced transparency (EIT) cooling scheme for a cantilever with a nitrogen-vacancy center attached in a non-uniform magnetic field using dynamical Zeeman effect. In our scheme, the Zeeman effect combined with the quantum interference effect enhances the desired cooling transition and suppresses the undesired heating transitions. As a result, the cantilever can be cooled down to nearly the vibrational ground state under realistic experimental conditions within a short time. This efficient optical EIT cooling scheme can be reduced to the typical EIT cooling scheme under special conditions. PMID:24514521

  13. Hard X-ray Ptychography: Making It Cool, Colorful and Fast

    NASA Astrophysics Data System (ADS)

    Deng, Junjing

    Ptychography is a recently developed coherent imaging technique for extended objects, with a resolution not limited by the lens. Because X-rays have short wavelengths and high penetration ability, X-ray ptychography provides a powerful and unique tool for studying thick samples at high spatial resolution. We have advanced X-ray ptychography by making it cool, colorful, and fast. We make it cool by carrying out ptychography experiments at cryogenic conditions to image frozen-hydrated specimens. This largely removes the limitations of radiation damage on the achievable resolution, and allows one to obtain excellent preservation of structure and chemistry in biological specimens. We make it colorful by combining it with X-ray fluorescence measurements of chemical element distributions. In studies of biological specimens, this means that ptychography can reveal cellular ultrastructure at high contrast and at a resolution well beyond that of X-ray focusing optics, while X-ray fluorescence is used to simultaneously image the distribution of trace elements in cells (such as metals that play key roles in cell functions and which can be used in various disease therapeutic agents). Because X-ray fluorescence is not very sensitive for showing the light elements that comprise the majority of cellular materials, this combined approach provides the unique tool to obtain simultaneous views of ultrastructure and elemental compositions of specimens. We make it fast by using continuous-scan (or "fly-scan") methods. Conventional ptychography is implemented in a move-settle-measure approach, which is slow due to the positioning overheads. To overcome this bottleneck, we have developed fly-scan ptychography that is able to speed up the data collection, and real time on-site data analysis can be achieved by using a parallelized reconstruction code. With these advances, we conducted combined cryo X-ray ptychography and fluorescence imaging at 5.2 keV in a more practical way using fly

  14. Conceptual Design study of Small Long-life Gas Cooled Fast Reactor With Modified CANDLE Burn-up Scheme

    SciTech Connect

    Nur Asiah, A.; Su'ud, Zaki; Ferhat, A.; Sekimoto, H.

    2010-06-22

    In this paper, conceptual design study of Small Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. In this study Gas Cooled Fast Reactor is slightly modified by employing modified CANDLE burn-up scheme so that it can use Natural Uranium as fuel cycle input. Due to their hard spectrum, GCFR in this study showed very good performance in converting U-238 to plutonium in order to maintain the operation condition requirement of long-life reactors. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. With such condition we got an optimal design of 325 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input. The average discharge burn-up is about 290 GWd/ton HM.

  15. Conceptual Design study of Small Long-life Gas Cooled Fast Reactor With Modified CANDLE Burn-up Scheme

    NASA Astrophysics Data System (ADS)

    Nur Asiah, A.; Su'ud, Zaki; Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this paper, conceptual design study of Small Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. In this study Gas Cooled Fast Reactor is slightly modified by employing modified CANDLE burn-up scheme so that it can use Natural Uranium as fuel cycle input. Due to their hard spectrum, GCFR in this study showed very good performance in converting U-238 to plutonium in order to maintain the operation condition requirement of long-life reactors. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. With such condition we got an optimal design of 325 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input. The average discharge burn-up is about 290 GWd/ton HM.

  16. Analysis and Development of A Robust Fuel for Gas-Cooled Fast Reactors

    SciTech Connect

    Knight, Travis W

    2010-01-31

    The focus of this effort was on the development of an advanced fuel for gas-cooled fast reactor (GFR) applications. This composite design is based on carbide fuel kernels dispersed in a ZrC matrix. The choice of ZrC is based on its high temperature properties and good thermal conductivity and improved retention of fission products to temperatures beyond that of traditional SiC based coated particle fuels. A key component of this study was the development and understanding of advanced fabrication techniques for GFR fuels that have potential to reduce minor actinide (MA) losses during fabrication owing to their higher vapor pressures and greater volatility. The major accomplishments of this work were the study of combustion synthesis methods for fabrication of the ZrC matrix, fabrication of high density UC electrodes for use in the rotating electrode process, production of UC particles by rotating electrode method, integration of UC kernels in the ZrC matrix, and the full characterization of each component. Major accomplishments in the near-term have been the greater characterization of the UC kernels produced by the rotating electrode method and their condition following the integration in the composite (ZrC matrix) following the short time but high temperature combustion synthesis process. This work has generated four journal publications, one conference proceeding paper, and one additional journal paper submitted for publication (under review). The greater significance of the work can be understood in that it achieved an objective of the DOE Generation IV (GenIV) roadmap for GFR Fuel—namely the demonstration of a composite carbide fuel with 30% volume fuel. This near-term accomplishment is even more significant given the expected or possible time frame for implementation of the GFR in the years 2030 -2050 or beyond.

  17. The Gas-Cooled Fast Reactor: Report on Safety System Design for Decay Heat Removal

    SciTech Connect

    K. D. Weaver; T. Marshall; T. Y. C. Wei; E. E. Feldman; M. J. Driscoll; H. Ludewig

    2003-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radiotoxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. This report addresses/discusses the decay heat removal options available to the GFR, and the current solutions. While it is possible to design a GFR with complete passive safety (i.e., reliance solely on conductive and radiative heat transfer for decay heat removal), it has been shown that the low power density results in unacceptable fuel cycle costs for the GFR. However, increasing power density results in higher decay heat rates, and the attendant temperature increase in the fuel and core. Use of active movers, or blowers/fans, is possible during accident conditions, which only requires 3% of nominal flow to remove the decay heat. Unfortunately, this requires reliance on active systems. In order to incorporate passive systems, innovative designs have been studied, and a mix of passive and active systems appears to meet the requirements for decay heat removal during accident conditions.

  18. Investigations of Alternative Steam Generator Location and Flatter Core Geometry for Lead-Cooled Fast Reactors

    SciTech Connect

    Carlsson, Johan; Tucek, Kamil; Wider, Hartmut

    2006-07-01

    This paper concerns two independent safety investigations on critical and sub-critical heavy liquid metal cooled fast reactors using simple flow paths. The first investigation applies to locating the steam generators in the risers instead of the down-comers of a simple flow path designed sub-critical reactor of 600 MW{sub th} power. This was compared to a similar design, but with the steam generators located in the downcomers. The transients investigated were Total-Loss-of-Power and unprotected Loss-Of-Flow. It is shown that this reactor peaks at 1041 K after 29 hours during a Total-Loss-Of-Power accident. The difference between locating the steam generators in the risers and the downcomers is insignificant for this accident type. During an unprotected Loss-Of-Flow accident at full power, the core outlet temperature stabilizes at 1010 K, which is 337 K above nominal outlet temperature. The second investigation concerns a 1426 MW{sub th} critical reactor where the influence of the core height versus the core outlet temperature is studied during an unprotected Loss-Of-Flow and Total-Loss-Of-Power accident. A pancake type core geometry of 1.0 m height and 5.8 m diameter, is compared to a compact core of 2 m height and 4.5 m diameter. Moderators, like BeO and hydrides, and their influence on safety coefficients and burnup swings are also presented. Both cores incinerate transuranics from spent LWR fuel with minor actinide fraction of 5%. We show that LFRs can be designed both to breed and burn transuranics from LWRs. It is shown that the hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. The computational fluid dynamics code STAR-CD was used for all thermal hydraulic calculations, and the MCNP and MCB for neutronics, and burn-up calculations. (authors)

  19. Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

    SciTech Connect

    Gordeev, S.; Hering, W.; Schikorr, M.; Stieglitz, R.

    2012-07-01

    CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

  20. RELAP5 Analysis of the Hybrid Loop-Pool Design for Sodium Cooled Fast Reactors

    SciTech Connect

    Hongbin Zhang; Haihua Zhao; Cliff Davis

    2008-06-01

    An innovative hybrid loop-pool design for sodium cooled fast reactors (SFR-Hybrid) has been recently proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to improve economics and safety of SFRs. In the hybrid loop-pool design, primary loops are formed by connecting the reactor outlet plenum (hot pool), intermediate heat exchangers (IHX), primary pumps and the reactor inlet plenum with pipes. The primary loops are immersed in the cold pool (buffer pool). Passive safety systems -- modular Pool Reactor Auxiliary Cooling Systems (PRACS) – are added to transfer decay heat from the primary system to the buffer pool during loss of forced circulation (LOFC) transients. The primary systems and the buffer pool are thermally coupled by the PRACS, which is composed of PRACS heat exchangers (PHX), fluidic diodes and connecting pipes. Fluidic diodes are simple, passive devices that provide large flow resistance in one direction and small flow resistance in reverse direction. Direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) are immersed in the cold pool to transfer decay heat to the environment by natural circulation. To prove the design concepts, especially how the passive safety systems behave during transients such as LOFC with scram, a RELAP5-3D model for the hybrid loop-pool design was developed. The simulations were done for both steady-state and transient conditions. This paper presents the details of RELAP5-3D analysis as well as the calculated thermal response during LOFC with scram. The 250 MW thermal power conventional pool type design of GNEP’s Advanced Burner Test Reactor (ABTR) developed by Argonne National Laboratory was used as the reference reactor core and primary loop design. The reactor inlet temperature is 355 °C and the outlet temperature is 510 °C. The core design is the same as that for ABTR. The steady state buffer pool temperature is the same as the reactor inlet

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

    NASA Astrophysics Data System (ADS)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-01

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

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

    SciTech Connect

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-31

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

  3. Effect of cold rolling on the oxidation resistance of T91 steel in oxygen-saturated stagnant liquid lead-bismuth eutectic at 450 °C and 550 °C

    NASA Astrophysics Data System (ADS)

    Dong, Hong; Ye, Zhongfei; Wang, Pei; Li, Dianzhong; Zhang, Yutuo; Li, Yiyi

    2016-08-01

    The compatibility of T91 steels having different preparation processes with oxygen-saturated stagnant lead-bismuth eutectic have been investigated at 450 °C and 550 °C. It is found that cold rolling decreases the thickness of the oxide scale of T91 steel by forming a continuous enhanced Cr-rich belt in the inner oxide layer next to the internal oxidation zone, which is attributed to the rapid diffusion of Cr induced by numerous non-equilibrium grain boundaries and migrating dislocations.

  4. Capturing high temperature protein conformations for low-temperature study using ultra-fast cooling

    NASA Astrophysics Data System (ADS)

    Moreau, David; Atakisi, Hakan; Thorne, Robert

    protocols for cooling biomolecular crystals for x-ray cryocrystallography are poorly controlled, leading to crystal-to-crystal and within-crystal non-isomorphism. Furthermore, cooling times below the protein-solvent glass transition of .1 s provide ample time for biological temperature conformations to depopulate and shift. To address these issues, methods and apparatus for cooling biomolecular crystals at rates approaching 100,000 K/s have been developed. These cooling rates are sufficient to eliminate ice formation on cooling without use of cryoprotectants, and to quench additional high-temperature conformations for low-temperature study. Time scales for conformational relaxation can be characterized using variable cooling rates. Possible extension of these methods to maximize conformational quenching will be discussed.

  5. Simulation of Radioactive Corrosion Product in Primary Cooling System of Japanese Sodium-Cooled Fast Breeder Reactor

    NASA Astrophysics Data System (ADS)

    Matuo, Youichirou; Miyahara, Shinya; Izumi, Yoshinobu

    Radioactive Corrosion Product (CP) is a main cause of personal radiation exposure during maintenance with no breached fuel in fast breeder reactor (FBR) plants. The most important CP is 54Mn and 60Co. In order to establish techniques of radiation dose estimation for radiation workers in radiation-controlled areas of the FBR, the PSYCHE (Program SYstem for Corrosion Hazard Evaluation) code was developed. We add the Particle Model to the conventional PSYCHE analytical model. In this paper, we performed calculation of CP transfer in JOYO using an improved calculation code in which the Particle Model was added to the PSYCHE. The C/E (calculated / experimentally observed) value for CP deposition was improved through use of this improved PSYCHE incorporating the Particle Model. Moreover, among the percentage of total radioactive deposition accounted for by CP in particle form, 54Mn was estimated to constitute approximately 20 % and 60Co approximately 40 % in the cold-leg region. These calculation results are consistent with the measured results for the actual cold-leg piping in the JOYO.

  6. New Fuel Cycle and Fuel Management Options in Heavy Liquid Metal-Cooled Reactors

    SciTech Connect

    Greenspan, Ehud; Hejzlar, Pavel; Sekimoto, Hiroshi; Toshinsky, Georgy; Wade, David

    2005-08-15

    Fast reactors cooled by lead or lead-bismuth alloy offer new interesting fuel cycle and fuel management options by virtue of the superb neutronics and safety features of these heavy liquid metal (HLM) coolants. One option is once-for-life cores having relatively low power density. These cores are fueled in the factory; there is no refueling or fuel shuffling on site. A second option is very long-life cores being made of a fissioning zone and a natural uranium blanket zone. The fissioning zone very slowly drifts toward the blanket. A third option is multirecycling of light water reactor (LWR) discharged fuel without partitioning of transuranics (TRUs) in fuel-self-sustaining reactors. LWR spent fuel could provide the initial fuel loading after extracting fission products and {approx}90% of its uranium. The makeup fuel is natural or depleted uranium. A fourth option is the high-burnup once-through fuel cycle using natural or depleted uranium feed. The initial fuel loading of this reactor is a mixture of enriched and natural uranium. The natural uranium utilization is 10 to 20 times higher than that of a once-through LWR. A fifth option is transmutation of TRUs from LWRs using critical HLM-cooled reactors; such reactors could be designed to have the same high actinide burning capability of accelerator-driven systems and have comparable safety, but at a substantially lower cost. These novel reactor designs and fuel management options are hereby reviewed.

  7. Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

    SciTech Connect

    Perret, G.; Pattupara, R. M.; Girardin, G.; Chawla, R.

    2012-07-01

    The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

  8. Gas-cooled fast breeder reactor. Quarterly progress report, February 1-April 30, 1980

    SciTech Connect

    Not Available

    1980-05-01

    Information is presented concerning the reactor vessel; reactivity control mechanisms and instrumentation; reactor internals; primary coolant circuits;core auxiliary cooling system; reactor core; systems engineering; and reactor safety and reliability;

  9. Stability analysis of a natural circulation lead-cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Lu, Qiyue

    This dissertation is aimed at nuclear-coupled thermal hydraulics stability analysis of a natural circulation lead cooled fast reactor design. The stability concerns arise from the fact that natural circulation operation makes the system susceptible to flow instabilities similar to those observed in boiling water reactors. In order to capture the regional effects, modal expansion method which incorporates higher azimuthal modes is used to model the neutronics part of the system. A reduced order model is used in this work for the thermal-hydraulics. Consistent with the number of heat exchangers (HXs), the reactor core is divided into four equal quadrants. Each quadrant has its corresponding external segments such as riser, plenum, pipes and HX forming an equivalent 1-D closed loop. The local pressure loss along the loop is represented by a lumped friction factor. The heat transfer process in the HX is represented by a model for the coolant temperature at the core inlet that depends on the coolant temperature at the core outlet and the coolant velocity. Additionally, time lag effects are incorporated into this HX model due to the finite coolant speed. A conventional model is used for the fuel pin heat conduction to couple the neutronics and thermal-hydraulics. The feedback mechanisms include Doppler, axial/radial thermal expansion and coolant density effects. These effects are represented by a linear variation of the macroscopic cross sections with the fuel temperature. The weighted residual method is used to convert the governing PDEs to ODEs. Retaining the first and second modes, leads to six ODEs for neutronics, and five ODEs for the thermal-hydraulics in each quadrant. Three models are developed. These are: 1) natural circulation model with a closed coolant flow path but without coupled neutronics, 2) forced circulation model with constant external pressure drop across the heated channels but without coupled neutronics, 3) coupled system including neutronics with

  10. Fast 4 to 8 GHz Debuncher betatron stochastic cooling for the Tevatron upgrade

    SciTech Connect

    Mtingwa, S.K.

    1987-07-01

    The present 2 to 4 Ghz Debuncher betatron stochastic cooling systems are designed to cool 7 x 10/sup 7/ antiprotons with initial transverse emittances of 20..pi.. mm-mrad down to 7..pi.. mm-mrad in 2 seconds. For the Tevatron upgrade, it would be advantageous to cool twice the number of antiprotons from 20..pi.. mm-mrad to 1..pi..d mm-mrad in half the time. We explore the possibility of achieving this goal with 4 to 8 Ghz stochastic cooling systems. Since the transverse beam size is initially more than twice the width of the cooling electrodes, we compare the effectiveness of one, two, and three plate pairs spaced strategically along the transverse beam direction. We also compare constant gap size electrodes with that for gap sizes which track the beam size as the emittance is reduced. We find that the three plate pair variable gap geometry performs the best, closely followed by the two plate pair variable gap geometry. Because of lower production costs we would suggest use of the latter. 8 refs., 7 figs., 37 tabs.

  11. Investigation of Nuclear Data Libraries with TRIPOLI-4 Monte Carlo Code for Sodium-cooled Fast Reactors

    NASA Astrophysics Data System (ADS)

    Lee, Y.-K.; Brun, E.

    2014-04-01

    The Sodium-cooled fast neutron reactor ASTRID is currently under design and development in France. Traditional ECCO/ERANOS fast reactor code system used for ASTRID core design calculations relies on multi-group JEFF-3.1.1 data library. To gauge the use of ENDF/B-VII.0 and JEFF-3.1.1 nuclear data libraries in the fast reactor applications, two recent OECD/NEA computational benchmarks specified by Argonne National Laboratory were calculated. Using the continuous-energy TRIPOLI-4 Monte Carlo transport code, both ABR-1000 MWth MOX core and metallic (U-Pu) core were investigated. Under two different fast neutron spectra and two data libraries, ENDF/B-VII.0 and JEFF-3.1.1, reactivity impact studies were performed. Using JEFF-3.1.1 library under the BOEC (Beginning of equilibrium cycle) condition, high reactivity effects of 808 ± 17 pcm and 1208 ± 17 pcm were observed for ABR-1000 MOX core and metallic core respectively. To analyze the causes of these differences in reactivity, several TRIPOLI-4 runs using mixed data libraries feature allow us to identify the nuclides and the nuclear data accounting for the major part of the observed reactivity discrepancies.

  12. Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

    SciTech Connect

    Meriyanti; Su'ud, Zaki; Rijal, K.; Zuhair; Ferhat, A.; Sekimoto, H.

    2010-06-22

    In this study a feasibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850 deg. C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticality was obtained for this reactor.

  13. Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

    NASA Astrophysics Data System (ADS)

    Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

  14. Safety and core design of large liquid-metal cooled fast breeder reactors

    NASA Astrophysics Data System (ADS)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  15. Ferritic steels for sodium-cooled fast reactors: Design principles and challenges

    NASA Astrophysics Data System (ADS)

    Raj, Baldev; Vijayalakshmi, M.

    2010-09-01

    An overview of the current status of development of ferritic steels for emerging fast reactor technologies is presented in this paper. The creep-resistant 9-12Cr ferritic/martensitic steels are classically known for steam generator applications. The excellent void swelling resistance of ferritic steels enabled the identification of their potential for core component applications of fast reactors. Since then, an extensive knowledge base has been generated by identifying the empirical correlations between chemistry of the steels, heat treatment, structure, and properties, in addition to their in-reactor behavior. A few concerns have also been identified which pertain to high-temperature irradiation creep, embrittlement, Type IV cracking in creep-loaded weldments, and hard zone formation in dissimilar joints. The origin of these problems and the methodologies to overcome the limitations are highlighted. Finally, the suitability of the ferritic steels is re-evaluated in the emerging scenario of the fast reactor technology, with a target of achieving better breeding ratio and improved thermal efficiency.

  16. Behavior of fast moving flow of compressible gas in cylindrical pipe in presence of cooling

    NASA Technical Reports Server (NTRS)

    Varshavsky, G A

    1951-01-01

    For compressible flow with friction in a cylindrical pipe the momentum, continuity, and heat-transfer equations are examined to determine whether an increase in Mach number ("thermal" Laval nozzle) is obtainable through heat conduction from the gas through the pipe walls. The analysis is based on the assumption that the wall temperature is negligibly small in comparison with the stagnation temperature of the gas. The analysis leads to a negative result. When the gas cooling is increased by also considering radiation to the wall, a limited region at high temperatures is obtained where Mach number increases were theoretically possible. Obtaining this condition practically is considered impossible.

  17. Ways of improvement for the materials of sodium cooled fast reactors

    SciTech Connect

    Horowitz, E.

    2012-07-01

    The French sodium cooled prototype reactor ASTRID will take into account 'Generation IV' requirements, especially a long operational life-time (60 years) and a high efficiency. The good behavior of austenitic steel AISI316L(N), should be confirmed for a use, in moderately irradiated and unirradiated parts of ASTRID. Parts recovered from dismantled French sodium-cooled reactors will be characterized. Further experiments must be carried out concerning ageing of these components. Other materials will be chosen for fuel wrapping and cladding, in order to reduce creep and swelling under irradiation, (either conventional, or oxide-dispersed strengthened steels (ODSS). Corrosion of ODSS in the presence of sodium needs a serious assessment The lifetime of primary pumps components made of Duplex steels should also be assessed. The disruptions in steam generator tubes should be minimized and controlled; therefore, optimised designs and geometries must be established before defining the corresponding materials. Either Modified 9Cr1Mo or Incoloy 800H, might be candidates;it will be necessary to check whether austenitic steels are compatible with Modified 9Cr1Mo or Incoloy 800H in the same circuit. For all materials, the best manufacturing processes must be combined with thermal, mechanical treatments; calculations of phase diagrams (CALPHAD) might be used to optimise both treatments and chemical compositions. (authors)

  18. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    SciTech Connect

    Su'ud, Zaki; Sekimoto, H.

    2014-09-30

    Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.

  19. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    NASA Astrophysics Data System (ADS)

    Su'ud, Zaki; Sekimoto, H.

    2014-09-01

    Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.

  20. Application of GRS method to evaluation of uncertainties of calculation parameters of perspective sodium-cooled fast reactor

    SciTech Connect

    Peregudov, A.; Andrianova, O.; Raskach, K.; Tsibulya, A.

    2012-07-01

    A number of recent studies have been devoted to the estimation of errors of reactor calculation parameters by the GRS (Generation Random Sampled) method. This method is based on direct sampling input data resulting in formation of random sets of input parameters which are used for multiple calculations. Once these calculations are performed, statistical processing of the calculation results is carried out to determine the mean value and the variance of each calculation parameter of interest. In our study this method is used for estimation of errors of calculation parameters (K{sub eff}, power density, dose rate) of a perspective sodium-cooled fast reactor. Neutron transport calculations were performed by the nodal diffusion code TRIGEX and Monte Carlo code MMK. (authors)

  1. An Analysis of Methanol and Hydrogen Production via High-Temperature Electrolysis Using the Sodium Cooled Advanced Fast Reactor

    SciTech Connect

    Shannon M. Bragg-Sitton; Richard D. Boardman; Robert S. Cherry; Wesley R. Deason; Michael G. McKellar

    2014-03-01

    Integration of an advanced, sodium-cooled fast spectrum reactor into nuclear hybrid energy system (NHES) architectures is the focus of the present study. A techno-economic evaluation of several conceptual system designs was performed for the integration of a sodium-cooled Advanced Fast Reactor (AFR) with the electric grid in conjunction with wind-generated electricity. Cases in which excess thermal and electrical energy would be reapportioned within an integrated energy system to a chemical plant are presented. The process applications evaluated include hydrogen production via high temperature steam electrolysis and methanol production via steam methane reforming to produce carbon monoxide and hydrogen which feed a methanol synthesis reactor. Three power cycles were considered for integration with the AFR, including subcritical and supercritical Rankine cycles and a modified supercritical carbon dioxide modified Brayton cycle. The thermal efficiencies of all of the modeled power conversions units were greater than 40%. A thermal efficiency of 42% was adopted in economic studies because two of the cycles either performed at that level or could potentially do so (subcritical Rankine and S-CO2 Brayton). Each of the evaluated hybrid architectures would be technically feasible but would demonstrate a different internal rate of return (IRR) as a function of multiple parameters; all evaluated configurations showed a positive IRR. As expected, integration of an AFR with a chemical plant increases the IRR when “must-take” wind-generated electricity is added to the energy system. Additional dynamic system analyses are recommended to draw detailed conclusions on the feasibility and economic benefits associated with AFR-hybrid energy system operation.

  2. Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

    SciTech Connect

    Jean Ragusa; Andrew Siegel; Jean-Michel Ruggieri

    2010-09-28

    The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.

  3. Studies of the Deteriorated Turbulent Heat Transfer Regime for the Gas-Cooled Fast Reactor Decay Heat Removal System

    SciTech Connect

    Jeong Ik Lee; Hejzlar, Pavel; Kazimi, Mujid S.; Saha, Pradip

    2006-07-01

    Increased reliance on passive emergency cooling using natural circulation of gas at elevated pressure is one of the major goals for the Gas-cooled Fast Reactor (GFR). Since GFR cores have high power density and low thermal inertia, the decay heat removal (DHR) in depressurization accidents is a key challenge. Furthermore, due to its high surface heat flux and low velocities under natural circulation in any post-LOCA scenario, three effects impair the capability of turbulent gas flow to remove heat from the GFR core, namely: (1) Acceleration effect (2) Buoyancy effect (3) Properties variation. This paper reviews previous work on heat transfer mechanisms and flow characteristics of the Deteriorated Turbulent Heat Transfer (DTHT) regime. It is shown that the GFR's DHR system has a potential for operating in the DTHT regime by performing a simple analysis. A description of the MIT/INL experimental facility designed and built to investigate the DTHT regime is provided together with the first test results. The first runs were performed in the forced convection regime to verify facility operation against well-established forced convection correlations. The results of the three runs at Reynolds numbers 6700, 8000 and 12800 showed good agreement with the Gnielinsky correlation [4], which is considered the best available heat transfer correlation in the forced convection regime and is valid for a large range of Reynolds and Prandtl numbers. However, even in the forced convection regime, the effect of heat transfer properties variation of the fluid was found to be still significant. (authors)

  4. Building on knowledge base of sodium cooled fast spectrum reactors to develop materials technology for fusion reactors

    NASA Astrophysics Data System (ADS)

    Raj, Baldev; Rao, K. Bhanu Sankara

    2009-04-01

    The alloys 316L(N) and Mod. 9Cr-1Mo steel are the major structural materials for fabrication of structural components in sodium cooled fast reactors (SFRs). Various factors influencing the mechanical behaviour of these alloys and different modes of deformation and failure in SFR systems, their analysis and the simulated tests performed on components for assessment of structural integrity and the applicability of RCC-MR code for the design and validation of components are highlighted. The procedures followed for optimal design of die and punch for the near net shape forming of petals of main vessel of 500 MWe prototype fast breeder reactor (PFBR); the safe temperature and strain rate domains established using dynamic materials model for forming of 316L(N) and 9Cr-1Mo steels components by various industrial processes are illustrated. Weldability problems associated with 316L(N) and Mo. 9Cr-1Mo are briefly discussed. The utilization of artificial neural network models for prediction of creep rupture life and delta-ferrite in austenitic stainless steel welds is described. The usage of non-destructive examination techniques in characterization of deformation, fracture and various microstructural features in SFR materials is briefly discussed. Most of the experience gained on SFR systems could be utilized in developing science and technology for fusion reactors. Summary of the current status of knowledge on various aspects of fission and fusion systems with emphasis on cross fertilization of research is presented.

  5. Study on In-Service Inspection Program and Inspection Technologies for Commercialized Sodium-Cooled Fast Reactor

    SciTech Connect

    Masato Ando; Shigenobu Kubo; Yoshio Kamishima; Toru Iitsuka

    2006-07-01

    The objective of in-service inspection of a nuclear power plant is to confirm integrity of function of components necessary to safety, and satisfy the needs to protect plant investment and to achieve high plant ability. The sodium-cooled fast reactor, which is designed in the feasibility study on commercialized fast reactor cycle systems in Japan, has two characteristics related to in-service inspection. The first is that all sodium coolant boundary structures have double-wall system. Continuous monitoring of the sodium coolant boundary structures are adopted for inspection. The second characteristic is the steam generator with double-wall-tubes. Volumetric testing is adopted to make sure that one of the tubes can maintain the boundary function in case of the other tube failure. A rational in-service inspection concept was developed taking these features into account. The inspection technologies were developed to implement in-service inspection plan. The under-sodium viewing system consisted of multi ultrasonic scanning transducers, which was used for imaging under-sodium structures. The under-sodium viewing system was mounted on the under-sodium vehicle and delivered to core internals. The prototype of under-sodium viewing system and vehicle were fabricated and performance tests were carried out under water. The laboratory experiments of volumetric testing for double-wall-tubes of steam generator, such as ultrasonic testing and remote-field eddy current testing, were performed and technical feasibility was assessed. (authors)

  6. Minor Actinide Recycle in Sodium Cooled Fast Reactors Using Heterogeneous Targets

    SciTech Connect

    Samuel Bays; Pavel Medvedev; Michael Pope; Rodolfo Ferrer; Benoit Forget; Mehdi Asgari

    2009-04-01

    This paper investigates the plausible design of transmutation target assemblies for minor actinides (MA) in Sodium Fast Reactors (SFR). A heterogeneous recycling strategy is investigated, whereby after each reactor pass, un-burned MAs from the targets are blended with MAs produced by the driver fuel and additional MAs from Spent Nuclear Fuel (SNF). A design iteration methodology was adopted for customizing the core design, target assembly design and matrix composition design. The overall design was constrained against allowable peak or maximum in-core performances. While respecting these criteria, the overall design was adjusted to reduce the total number of assemblies fabricated per refueling cycle. It was found that an inert metal-hydride MA-Zr-Hx target matrix gave the highest transmutation efficiency, thus allowing for the least number of targets to be fabricated per reactor cycle.

  7. Particle-bed gas-cooled fast reactor (PB-GCFR) design. Project final technical report (Sept 2001 - Aug 2003).

    SciTech Connect

    Taiwo, T. A.; Wei, T. Y. C.; Feldman, E. E.; Hoffman, E. A.; Fatone, M.; Holland, J. W.; Prokofiev, I. G.; Yang, W. S.; Palmiotti, G.; Hill, R. N.; Todosow, M.; Salvatores, M.; Gandini, A.

    2003-10-27

    The objective of this project is to develop a conceptual design of a particle-bed, gas-cooled fast reactor (PB-GCFR) core that meets the advanced reactor concept and enhanced proliferation-resistant goals of the US Department of Energy's NERI program. The key innovation of this project is the application of a fast neutron spectrum environment to enhance both the passive safety and transmutation characteristics of the advanced particle-bed and pebble-bed reactor designs. The PB-GCFR design is expected to produce a high-efficiency system with a low unit cost. It is anticipated that the fast neutron spectrum would permit small-sized units ({approx} 150 MWe) that can be built quickly and packaged into modular units, and whose production can be readily expanded as the demand grows. Such a system could be deployed globally. The goals of this two-year project are as follows: (1) design a reactor core that meets the future needs of the nuclear industry, by being passively safe with reduced need for engineered safety systems. This will entail an innovative core design incorporating new fuel form and type; (2) employ a proliferation-resistant fuel design and fuel cycle. This will be supported by a long-life core design that is refueled infrequently, and hence, reduces the potential for fuel diversion; (3) incorporate design features that permit use of the system as an efficient transmuter that could be employed for burning separated plutonium fuel or recycled LWR transuranic fuel, should the need arise; and (4) evaluate the fuel cycle for waste minimization and for the possibility of direct fuel disposal. The application of particle-bed fuel provides the promise of extremely high burnup and fission-product protection barriers that may permit direct disposal.

  8. “Universal” vitrification of cells by ultra-fast cooling

    PubMed Central

    Heo, Yun Seok; Nagrath, Sunitha; Moore, Alessandra L.; Zeinali, Mahnaz; Irimia, Daniel; Stott, Shannon L.; Toth, Thomas L.

    2015-01-01

    Long-term preservation of live cells is critical for a broad range of clinical and research applications. With the increasing diversity of cells that need to be preserved (e.g. oocytes, stem and other primary cells, genetically modified cells), careful optimization of preservation protocols becomes tedious and poses significant limitations for all but the most expert users. To address the challenge of long-term storage of critical, heterogeneous cell types, we propose a universal protocol for cell vitrification that is independent of cell phenotype and uses only low concentrations of cryoprotectant (1.5 M PROH and 0.5 M trehalose). We employed industrial grade microcapillaries made of highly conductive fused silica, which are commonly used for analytical chemistry applications. The minimal mass and thermal inertia of the microcapillaries enabled us to achieve ultrafast cooling rates up to 4,000 K/s. Using the same low, non-toxic concentration of cryoprotectant, we demonstrate high recovery and viability rates after vitrification for human mammary epithelial cells, rat hepatocytes, tumor cells from pleural effusions, and multiple cancer cell lines. PMID:25914896

  9. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

    SciTech Connect

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-01-01

    The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

  10. Performance of low smeared density sodium-cooled fast reactor metal fuel

    DOE PAGESBeta

    Porter, D. L.; H. J. M. Chichester; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-06-17

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactormore » designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low metaling points and gaseous precursors (Cs and Rb). Lastly, a model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.« less

  11. Performance of low smeared density sodium-cooled fast reactor metal fuel

    SciTech Connect

    Porter, D. L.; H. J. M. Chichester; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-06-17

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low metaling points and gaseous precursors (Cs and Rb). Lastly, a model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  12. Performance of low smeared density sodium-cooled fast reactor metal fuel

    NASA Astrophysics Data System (ADS)

    Porter, D. L.; Chichester, H. J. M.; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-10-01

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at.% burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  13. Performance of Low Smeared Density Sodium-cooled Fast Reactor Metal Fuel

    SciTech Connect

    Porter, D. L.; H. J. M. Chichester; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-10-01

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low metaling points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  14. Design and Testing of D.C. Conduction Pump for Sodium Cooled Fast Reactor

    SciTech Connect

    Nashine, B.K.; Dash, S.K.; Gurumurthy, K.; Rajan, M.; Vaidyanathan, G.

    2006-07-01

    DC Conduction pump immersed in sodium forms a part of Failed Fuel Location Module (FFLM) of 500 MWe Fast Breeder Reactor (PFBR) currently under construction. FFLM housed in control plug of the reactor, is used to locate the failed fuel sub-assembly due to clad rupture in the fuel pin. The DC conduction pump sucks the sodium from the top of fuel sub-assemblies through the selector valve and pumps the sodium to hold up for detecting the presence of delayed neutrons. Presence of delayed neutron is the indication of failure in the sampled fuel sub-assembly. The DC Conduction Pump was chosen because of its low voltage operation (2 V) where argon/alumina ceramic can provide required electrical insulation even at operating temperature of 560 deg. C without much complication on the manufacturing front. Sampling of sodium from top of different sub-assemblies is achieved by operation of selector valve in-conjunction with the drive motor. FFLM requires the pump to be immersed in sodium pool at {approx} 560 deg. C located above the fuel sub-assemblies in the reactor. The Pump of 0.36 m{sup 3}/h capacity and developing 1.45 Kg/ cm{sup 2} pressure was designed, manufactured and tested. The DC Conduction Pump has a stainless steel duct filled with liquid sodium, which is to be pumped. The stainless steel duct is kept in magnetic field obtained by means of electromagnet. The electromagnet is made of soft iron and the coil made of copper conductor surrounds the yoke portion of electromagnet. The external DC source of 2000 Amps, 2 Volt is used to send current through sodium placed in the stainless steel duct and the same current is sent through copper coil of electromagnet for producing required magneto motive force, which in turn produces required magnetic field. The interaction of current in sodium (placed in stainless steel duct) and magnetic field produced by the electromagnet in the duct region produces pumping force in the sodium. Electromagnet, copper coil, stainless steel

  15. Temperature dependence of liquid metal embrittlement susceptibility of a modified 9Cr-1Mo steel under low cycle fatigue in lead-bismuth eutectic at 160-450 °C

    NASA Astrophysics Data System (ADS)

    Gong, Xing; Marmy, Pierre; Qin, Ling; Verlinden, Bert; Wevers, Martine; Seefeldt, Marc

    2016-01-01

    Low cycle fatigue properties of a 9Cr-1Mo ferritic-martensitic steel (T91) have been tested in a low oxygen concentration (LOC) lead-bismuth eutectic (LBE) environment and in vacuum at 160-450 °C. The results show a clear fatigue endurance "trough" in LOC LBE, while no such a strong temperature dependence of the fatigue endurance is observed when the steel is tested in vacuum. The fractographic observations by means of scanning electron microscopy (SEM) show that ductile microdimples are prevalent on the fracture surfaces of the specimens tested in vacuum, whereas the fracture surfaces produced in LOC LBE at all the temperatures are characterized by quasi-cleavage. Interestingly, using electron backscatter diffraction (EBSD), martensitic laths close to the fatigue crack walls or to the fracture surfaces of the specimens tested in vacuum are found to have transformed into very fine equiaxed subgrains. Nevertheless, such microstructural modifications do not happen to the specimens tested in LOC LBE at 160-450 °C. These interesting microstructural distinctions indicate that liquid metal embrittlement (LME) is able to occur throughout the fatigue crack propagation phase in the full range of the temperatures investigated, i.e. LME is not very sensitive to temperature during the fatigue crack propagation.

  16. The development of a realistic source term for sodium-cooled fast reactors : assessment of current status and future needs.

    SciTech Connect

    LaChance, Jeffrey L.; Phillips, Jesse; Parma, Edward J., Jr.; Olivier, Tara Jean; Middleton, Bobby D.

    2011-06-01

    Sodium-cooled fast reactors (SFRs) continue to be proposed and designed throughout the United States and the world. Although the number of SFRs actually operating has declined substantially since the 1980s, a significant interest in advancing these types of reactor systems remains. Of the many issues associated with the development and deployment of SFRs, one of high regulatory importance is the source term to be used in the siting of the reactor. A substantial amount of modeling and experimental work has been performed over the past four decades on accident analysis, sodium coolant behavior, and radionuclide release for SFRs. The objective of this report is to aid in determining the gaps and issues related to the development of a realistic, mechanistically derived source term for SFRs. This report will allow the reader to become familiar with the severe accident source term concept and gain a broad understanding of the current status of the models and experimental work. Further, this report will allow insight into future work, in terms of both model development and experimental validation, which is necessary in order to develop a realistic source term for SFRs.

  17. Reanalysis of the Gas-cooled fast reactor experiments at the zero power facility Proteus - Spectral indices

    NASA Astrophysics Data System (ADS)

    Perret, G.; Pattupara, R. M.; Girardin, G.; Chawla, R.

    2013-03-01

    PROTEUS is a zero power reactor at the Paul Scherrer Institute which has been employed during the 1970's to study experimentally the physics of the gas-cooled fast reactor. Reaction rate distributions, flux spectrum and reactivity effects have been measured in several configurations featuring PuO2/UO2 fuel, absorbers, large iron shields, and thorium oxide and thorium metal fuel either distributed quasihomogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. This papers focus on the spectral indices - including fission and capture in 232Th and 237Np - measured in the reference PuO2/UO2 lattices and their predictions with an MCNPX model specially developed for the PROTEUS-GCFR core. Predictions were obtained with JEFF-3.1 and -3.11, ENDF/B-VII.0 and VII.1, and JENDL-3.3 and -4.0. A general good agreement was demonstrated. The ratio of 232Th fission to 239Pu fission, however, was under-predicted by 8.7±2.1% and 6.5±2.1% using ENDF/B-VII.0 and VII.1, respectively. Finally, the capture rates in 237Np tended to be underpredicted by the JEFF and JENDL libraries, although the new cross section in JEFF-3.1.1 slightly improved the 237Np capture to 239Pu fission results (3.4±2.4%).

  18. Development of Advanced 9Cr Ferritic-Martensitic Steels and Austenitic Stainless Steels for Sodium-Cooled Fast Reactor

    SciTech Connect

    Sham, Sam; Tan, Lizhen; Yamamoto, Yukinori

    2013-01-01

    Ferritic-martensitic (FM) steel Grade 92, with or without thermomechanical treatment (TMT), and austenitic stainless steels HT-UPS (high-temperature ultrafine precipitate strengthening) and NF709 were selected as potential candidate structural materials in the U.S. Sodium-cooled Fast Reactor (SFR) program. The objective is to develop advanced steels with improved properties as compared with reference materials such as Grade 91 and Type 316H steels that are currently in nuclear design codes. Composition modification and/or processing optimization (e.g., TMT and cold-work) were performed to improve properties such as resistance to thermal aging, creep, creep-fatigue, fracture, and sodium corrosion. Testings to characterize these properties for the advanced steels were conducted by the Idaho National Laboratory, the Argonne National Laboratory and the Oak Ridge National Laboratory under the U.S. SFR program. This paper focuses on the resistance to thermal aging and creep of the advanced steels. The advanced steels exhibited up to two orders of magnitude increase in creep life compared to the reference materials. Preliminary results on the weldment performance of the advanced steels are also presented. The superior performance of the advanced steels would improve reactor design flexibility, safety margins and economics.

  19. Impact of Fission Products Impurity on the Plutonium Content of Metal- and Oxide- Fuels in Sodium Cooled Fast Reactors

    SciTech Connect

    Hikaru Hiruta; Gilles Youinou

    2013-09-01

    This short report presents the neutronic analysis to evaluate the impact of fission product impurity on the Pu content of Sodium-cooled Fast Reactor (SFR) metal- and oxide- fuel fabrication. The similar work has been previously done for PWR MOX fuel [1]. The analysis will be performed based on the assumption that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate SFR fuels. Only non-gaseous FPs have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1 of Reference 1). Throughout of this report, we define the mixture of Pu and FPs as PuFP. The main objective of this analysis is to quantify the increase of the Pu content of SFR fuels necessary to maintain the same average burnup at discharge independently of the amount of FP in the Pu stream, i.e. independently of the PuFP composition. The FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

  20. Generation IV nuclear energy system initiative. Large GFR core subassemblydesign for the Gas-Cooled Fast Reactor.

    SciTech Connect

    Hoffman, E. A.; Kulak, R. F.; Therios, I. U.; Wei, T. Y. C.

    2006-07-31

    Gas-cooled fast reactor (GFR) designs are being developed to meet Gen IV goals of sustainability, economics, safety and reliability, and proliferation resistance and physical protection as part of an International Generation IV Nuclear Energy System Research Initiative effort. Different organizations are involved in the development of a variety of GFR design concepts. The current analysis has focused on the evaluation of low-pressure drop, pin-core designs with favorable passive cooling properties. Initial evaluation of the passive cooling safety case for the GFR during depressurized decay heat removal accidents with concurrent loss of electric power have resulted in requirements for a reduction of core power density to the 100 w/cc level and a low core pressure drop of 0.5 bars. Additional design constraints and the implementation of their constraints are evaluated in this study to enhance and passive cooling properties of the reactor. Passive cooling is made easier by a flat radial distribution of the decay heat. One goal of this study was to evaluate the radial power distribution and determine to what extent it can be flattened, since the decay heat is nearly proportional to the fission power at shutdown. In line with this investigation of the radial power profile, an assessment was also made of the control rod configuration. The layout provided a large number of control rod locations with a fixed area provided for control rods. The number of control rods was consistent with other fast reactor designs. The adequacy of the available control rod locations was evaluated. Future studies will be needed to optimize the control rod designs and evaluate the shutdown system. The case for low pressure drop core can be improved by the minimization of pressure drop sources such as the number of required fuel spacers in the subassembly design and by the details of the fuel pin design. The fuel pin design is determined by a number of neutronic, thermal-hydraulic (gas dynamics

  1. Gas-Cooled Fast Breeder Reactor Preliminary Safety Information Document, Amendment 10. GCFR residual heat removal system criteria, design, and performance

    SciTech Connect

    Not Available

    1980-09-01

    This report presents a comprehensive set of safety design bases to support the conceptual design of the gas-cooled fast breeder reactor (GCFR) residual heat removal (RHR) systems. The report is structured to enable the Nuclear Regulatory Commission (NRC) to review and comment in the licensability of these design bases. This report also presents information concerning a specific plant design and its performance as an auxiliary part to assist the NRC in evaluating the safety design bases.

  2. Investigations on the heat transport capability of a cryogenic oscillating heat pipe and its application in achieving ultra-fast cooling rates for cell vitrification cryopreservation☆

    PubMed Central

    Han, Xu; Ma, Hongbin; Jiao, Anjun; Critser, John K.

    2010-01-01

    Theoretically, direct vitrification of cell suspensions with relatively low concentrations (~1 M) of permeating cryoprotective agents (CPA) is suitable for cryopreservation of almost all cell types and can be accomplished by ultra-fast cooling rates that are on the order of 106–7 K/min. However, the methods and devices currently available for cell cryopreservation cannot achieve such high cooling rates. In this study, we constructed a novel cryogenic oscillating heat pipe (COHP) using liquid nitrogen as its working fluid and investigated its heat transport capability to assess its application for achieving ultra-fast cooling rates for cell cryopreservation. The experimental results showed that the apparent heat transfer coefficient of the COHP can reach 2 × 105 W/m2·K, which is two orders of the magnitude higher than traditional heat pipes. Theoretical analyzes showed that the average local heat transfer coefficient in the thin film evaporation region of the COHP can reach 1.2 × 106 W/m2·K, which is approximately 103 times higher than that achievable with standard pool-boiling approaches. Based on these results, a novel device design applying the COHP and microfabrication techniques is proposed and its efficiency for cell vitrification is demonstrated through numerical simulation. The estimated average cooling rates achieved through this approach is 106–7 K/min, which is much faster than the currently available methods and sufficient for achieving vitrification with relatively low concentrations of CPA. PMID:18430413

  3. Thermal criteria to compare fast reactors coolants for the intermediate loop

    SciTech Connect

    Saez, Manuel; Rodriguez, Gilles

    2007-07-01

    Fast Breeder Reactors (FBR) are typically using a liquid metal as the primary coolant. Up to now, sodium is the referenced coolant for all large-scale FBR, but lead and sodium-potassium alloy have both also been used successfully for smaller rigs. The French Atomic Energy Commission (CEA) has an extensive experience and significant expertise in Sodium cooled Fast Reactors (SFR) over the past 40 years of R and D and feedback experiments. Some improvements are needed on the SFR to meet the Generation IV goals, and in particular the safety and the reliability through the intermediate loop coolant. As sodium reacts exo-thermically with air and water and to eliminate the drawback of the water-sodium interaction when a steam generator tube is ruptured, CEA is involved in a substantial effort in order to investigate the interest to use an alternative coolant than sodium in the intermediate loop. This paper presents the main thermal criteria to compare Fast Reactors coolants for the intermediate loop under natural and forced convection. Neutronics considerations are not taken into account for the intermediate loop coolant. Transport, transfer and energetic criteria are analysed in the field of turbulent flows. Criteria are applied to the following potential coolant candidates: sodium, lithium, tin, bismuth, lead, lead-bismuth alloy, lead-lithium alloy, gallium, indium, potassium and sodium-potassium alloy. According to this thermal analysis, the gallium as heat transfer agent for the intermediate loop is considered as a promising candidate. For the discussion of the applicability of the gallium as heat transfer agent for the intermediate loop, a limited thermal hydraulic pre-sizing of a steam generator is undertaken using simple engineering methods implemented in COPERNIC code, a CEA tool dedicated to reactor systems pre-sizing. (authors)

  4. Fuel Cycle System Analysis Implications of Sodium-Cooled Metal-Fueled Fast Reactor Transuranic Conversion Ratio

    SciTech Connect

    Steven J. Piet; Edward A. Hoffman; Samuel E. Bays; Gretchen E. Matthern; Jacob J. Jacobson; Ryan Clement; David W. Gerts

    2013-03-01

    If advanced fuel cycles are to include a large number of fast reactors (FRs), what should be the transuranic (TRU) conversion ratio (CR)? The nuclear energy era started with the assumption that they should be breeder reactors (CR > 1), but the full range of possible CRs eventually received attention. For example, during the recent U.S. Global Nuclear Energy Partnership program, the proposal was burner reactors (CR < 1). Yet, more recently, Massachusetts Institute of Technology's "Future of the Nuclear Fuel Cycle" proposed CR [approximately] 1. Meanwhile, the French company EDF remains focused on breeders. At least one of the reasons for the differences of approach is different fuel cycle objectives. To clarify matters, this paper analyzes the impact of TRU CR on many parameters relevant to fuel cycle systems and therefore spans a broad range of topic areas. The analyses are based on a FR physics parameter scan of TRU CR from 0 to [approximately]1.8 in a sodium-cooled metal-fueled FR (SMFR), in which the fuel from uranium-oxide-fueled light water reactors (LWRs) is recycled directly to FRs and FRs displace LWRs in the fleet. In this instance, the FRs are sodium cooled and metal fueled. Generally, it is assumed that all TRU elements are recycled, which maximizes uranium ore utilization for a given TRU CR and waste radiotoxicity reduction and is consistent with the assumption of used metal fuel separated by electrochemical means. In these analyses, the fuel burnup was constrained by imposing a neutron fluence limit to fuel cladding to the same constant value. This paper first presents static, time-independent measures of performance for the LWR [right arrow] FR fuel cycle, including mass, heat, gamma emission, radiotoxicity, and the two figures of merit for materials for weapon attractiveness developed by C. Bathke et al. No new fuel cycle will achieve a static equilibrium in the foreseeable future. Therefore, additional analyses are shown with dynamic, time

  5. Investigation of alternative layouts for the supercritical carbon dioxide Brayton cycle for a sodium-cooled fast reactor.

    SciTech Connect

    Moisseytsev, A.; Sienicki, J. J.

    2009-07-01

    Analyses of supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle performance have largely settled on the recompression supercritical cycle (or Feher cycle) incorporating a flow split between the main compressor downstream of heat rejection, a recompressing compressor providing direct compression without heat rejection, and high and low temperature recuperators to raise the effectiveness of recuperation and the cycle efficiency. Alternative cycle layouts have been previously examined by Angelino (Politecnico, Milan), by MIT (Dostal, Hejzlar, and Driscoll), and possibly others but not for sodium-cooled fast reactors (SFRs) operating at relatively low core outlet temperature. Thus, the present authors could not be sure that the recompression cycle is an optimal arrangement for application to the SFR. To ensure that an advantageous alternative layout has not been overlooked, several alternative cycle layouts have been investigated for a S-CO{sub 2} Brayton cycle coupled to the Advanced Burner Test Reactor (ABTR) SFR preconceptual design having a 510 C core outlet temperature and a 470 C turbine inlet temperature to determine if they provide any benefit in cycle performance (e.g., enhanced cycle efficiency). No such benefits were identified, consistent with the previous examinations, such that attention was devoted to optimizing the recompression supercritical cycle. The effects of optimizing the cycle minimum temperature and pressure are investigated including minimum temperatures and/or pressures below the critical values. It is found that improvements in the cycle efficiency of 1% or greater relative to previous analyses which arbitrarily fixed the minimum temperature and pressure can be realized through an optimal choice of the combination of the minimum cycle temperature and pressure (e.g., for a fixed minimum temperature there is an optimal minimum pressure). However, this leads to a requirement for a larger cooler for heat rejection which may impact the

  6. Highly c-axis-oriented monocrystalline Pb(Zr, Ti)O₃ thin films on si wafer prepared by fast cooling immediately after sputter deposition.

    PubMed

    Yoshida, Shinya; Hanzawa, Hiroaki; Wasa, Kiyotaka; Esashi, Masayoshi; Tanaka, Shuji

    2014-09-01

    We successfully developed sputter deposition technology to obtain a highly c-axis-oriented monocrystalline Pb(Zr, Ti)O3 (PZT) thin film on a Si wafer by fast cooling (~-180°C/min) of the substrate after deposition. The c-axis orientation ratio of a fast-cooled film was about 90%, whereas that of a slow-cooled (~-40°C/min) film was only 10%. The c-axis-oriented monocrystalline Pb(Zr0.5, Ti0.5)O3 films showed reasonably large piezoelectric coefficients, e(31,f) = ~-11 C/m(2), with remarkably small dielectric constants, ϵ(r) = ~220. As a result, an excellent figure of merit (FOM) was obtained for piezoelectric microelectromechanical systems (MEMS) such as a piezoelectric gyroscope. This c-axis orientation technology on Si will extend industrial applications of PZT-based thin films and contribute further to the development of piezoelectric MEMS. PMID:25167155

  7. Judd-Ofelt analysis, frequency upconversion, and infrared photoluminescence of Ho{sup 3+}-doped and Ho{sup 3+}/Yb{sup 3+}-codoped lead bismuth gallate oxide glasses

    SciTech Connect

    Zhou Bo; Pun, Edwin Yue-Bun; Yang Dianlai; Huang Lihui; Lin Hai

    2009-11-15

    Ho{sup 3+}-doped and Ho{sup 3+}/Yb{sup 3+}-codoped lead bismuth gallate (PBG) oxide glasses were prepared and their spectroscopic properties were investigated. The derived Judd-Ofelt intensity parameters (OMEGA{sub 2}=6.81x10{sup -20} cm{sup 2}, OMEGA{sub 4}=2.31x10{sup -20} cm{sup 2}, and OMEGA{sub 6}=0.67x10{sup -20} cm{sup 2}) indicate a higher asymmetry and stronger covalent environment for Ho{sup 3+} sites in PBG glass compared with those in tellurite, fluoride (ZBLAN), and some other lead-contained glasses. Intense frequency upconversion emissions peaking at 547, 662, and 756 nm as well as infrared emissions at 1.20 and 2.05 mum in Ho{sup 3+}/Yb{sup 3+}-codoped PBG glass were observed, confirming that energy transfer between Yb{sup 3+} and Ho{sup 3+} takes place, and a two-phonon-assisted energy transfer from Yb{sup 3+} to Ho{sup 3+} ions was determined by the calculation using phonon sideband theory. The 1.20 mum emission observed was primarily due to the weak multiphonon deexcitation originated from the small phonon energy of PBG glass (approx535 cm{sup -1}). A large product of emission cross-section and measured lifetime (9.93x10{sup -25} cm{sup 2} s) was obtained for the 1.20 mum emission and the gain coefficient dependence on wavelength with population inversion rate (P) was performed. The peak emission cross-section for 2.05 mum emission was calculated to be 4.75x10{sup -21} cm{sup 2}. The relative mechanism of Ho{sup 3+}-doped and Ho{sup 3+}/Yb{sup 3+}-codoped PBG glasses on their spectroscopic properties was also discussed. Our results suggest that Ho{sup 3+}/Yb{sup 3+}-doped PBG glasses are a good potential candidate for the frequency upconversion devices and infrared amplifiers/lasers.

  8. Feasibility study on ultralong-cycle operation and material performance for compact liquid metal-cooled fast reactors: a review work

    SciTech Connect

    Tak, Taewoo; Choe, Jiwon; Jeong, Yongjin; Lee, Deokjung; Kim, T. K.; Hong, Ser Gi

    2015-11-01

    This paper reviews the feasibility of ultralong-cycle operation on a compact liquid metal-cooled fast reactor (LMR) firstly by assessing the operation of a long-life fast reactor core and secondly by evaluating material performance in respect to both long-cycle operation and compact-size fast reactor. Many kinds of reactor concepts have been proposed, and LMR and small modular reactor (SMR) are the issued leading technologies for generation four (Gen-IV) reactor system development. The breed-and-burn strategy was proposed as a core burning strategy to operate a long cycle, and it has been evaluated in this paper with two reactor concepts: constant axial shape of neutron flux, nuclide densities, and power shape during life of energy and ultralong cycle fast reactor. In addition, Super-Safe, Small, and Simple and small modular fast reactor, compact LMR concepts, have been simulated to evaluate their long-life operation strategies. For the other practical issues, the materials for fuel, coolant, and structure have been identified and some of them are selected to have their performance optimized specifically for compact LMR with a long-cycle operation. It is believed that this comprehensive review will propose a proper direction for future reactor development and will be followed by the next step research for a complete reactor model with the other reactor components.

  9. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    NASA Astrophysics Data System (ADS)

    Monado, Fiber; Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Aziz, Ferhat; Permana, Sidik; Sekimoto, Hiroshi

    2014-02-01

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

  10. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    SciTech Connect

    Monado, Fiber; Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik; Aziz, Ferhat; Sekimoto, Hiroshi

    2014-02-12

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

  11. Thermal hydraulic analysis of advanced Pb-Bi cooled NPP using natural circulation

    NASA Astrophysics Data System (ADS)

    Novitrian, Su'ud, Zaki; Waris, Abdul

    2012-06-01

    We present thermal hydraulic analysis for a low power advanced nuclear reactor cooled by lead-bismuth eutectic. In this work is to study the thermal hydraulic analysis of a low power SPINNOR (Small Power Reactor, Indonesia, No On-site Refuelling) reactor with 125 MWth which a design a core with very small volume and fuel column height, resulting in a negative coolant temperature coefficient and very low channel pressure drop. And also at full power the heat can be completely removed by natural circulation in the primary circuit, thus eliminating the needs for pumps.

  12. Progress in the R and D Project on Oxide Dispersion Strengthened and Precipitation Hardened Ferritic Steels for Sodium Cooled Fast Breeder Reactor Fuels

    SciTech Connect

    Kaito, Takeji; Ohtsuka, Satoshi; Inoue, Masaki

    2007-07-01

    High burnup capability of sodium cooled fast breeder reactor (SFR) fuels depends significantly on irradiation performance of their component materials. Japan Atomic Energy Agency (JAEA) has been developing oxide dispersion strengthened (ODS) ferritic steels and a precipitation hardened (PH) ferritic steel as the most prospective materials for fuel pin cladding and duct tubes, respectively. Technology for small-scale manufacturing is already established, and several hundreds of ODS steel cladding tubes and dozens of PH steel duct tubes were successfully produced. We will step forward to develop manufacturing technology for mass production to supply these steels for future SFR fuels. Mechanical properties of the products were examined by out-of-pile and in-pile tests including material irradiation tests in the experimental fast reactor JOYO and foreign fast reactors. The material strength standards (MSSs) were tentatively compiled in 2005 for ODS steels and in 1993 for PH steel. In order to upgrade the MSSs and to demonstrate high burnup capability of the materials, we will perform a series of irradiation tests in BOR-60 and JOYO until 2015 and contribute to design study for a demonstration SFR of which operation is expected after 2025. (authors)

  13. Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

    SciTech Connect

    Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.; Wade, D. C.; Nikiforova, A.; Hanania, P.; Ryu, H. J.; Kulesza, K. P.; Kim, S. J.; Halsey, W. G.; Smith, C. F.; Brown, N. W.; Greenspan, E.; de Caro, M.; Li, N.; Hosemann, P.; Zhang, J.; Yu, H.; Nuclear Engineering Division; LLNL; LANL; Massachusetts Inst. of Tech.; Ecole des Mines de Paris; Oregon State Univ.; Univ.of California at Berkley

    2008-06-23

    This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been made at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics

  14. Formation of very hard electron and gamma-ray spectra of flat-spectrum radio quasars in the fast-cooling regime

    NASA Astrophysics Data System (ADS)

    Yan, Dahai; Zhang, Li; Zhang, Shuang-Nan

    2016-07-01

    In the external Compton scenario, we investigate the formation of a very hard electron spectrum in the fast-cooling regime, using a time-dependent emission model. It is shown that a very hard electron distribution, N^' }_e({γ ^' })∝ {γ ^' }^{-p}, with spectral index p ˜ 1.3 is formed below the minimum energy of injection electrons when inverse Compton scattering takes place in the Klein-Nishina regime, i.e. inverse Compton scattering of relativistic electrons on broad-line region radiation in flat-spectrum radio quasars. This produces a very hard gamma-ray spectrum and can explain in reasonable fashion the very hard Fermi-Large Area Telescope (LAT) spectrum of the flat-spectrum radio quasar 3C 279 during the extreme gamma-ray flare in 2013 December.

  15. Compact cryogenically cooled Ti:Sapphire dual multi-kilohertz amplifiers for synchrotron radiation ultra-fast x-ray applications

    SciTech Connect

    Feng, J.; Nasiatka, J.; Hertlein, M.; Rude, B.; Padmore, H.

    2013-05-15

    A titanium-doped sapphire regenerative dual-amplifier array operating at multi-kHz repetition rates has been developed for synchrotron radiation ultra-fast x-ray applications. The thermal lensing of the crystal in the amplifiers is virtually eliminated by cryogenic cooling of the laser crystal. The output energy of the amplifiers is measured to be greater than 2.6 mJ and the pulse length was compressed to less than 70 fs. The output laser mode is a near perfect Gaussian TEM00 with an M{sup 2} factor of 1.02. The performance of the amplifier system is in excellent agreement with theoretical calculation.

  16. Development of a multiphysics analysis system for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

    NASA Astrophysics Data System (ADS)

    Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

    2015-12-01

    A multiphysics analysis system for sodium-water reaction phenomena in a steam generator of sodium-cooled fast reactors was newly developed. The analysis system consists of the mechanistic numerical analysis codes, SERAPHIM, TACT, and RELAP5. The SERAPHIM code calculates the multicomponent multiphase flow and sodium-water chemical reaction caused by discharging of pressurized water vapor. Applicability of the SERAPHIM code was confirmed through the analyses of the experiment on water vapor discharging in liquid sodium. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The numerical models integrated into the TACT code were verified through some related experiments. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the tube rapidly heated by the reacting jet. The developed system enables evaluation of the wastage environment and the possibility of the failure propagation.

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

    SciTech Connect

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

    1981-04-01

    Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relate this profile to that generated by the coils in completed fuel pin simulators.

  18. Development of a multiphysics analysis system for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

    SciTech Connect

    Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

    2015-12-31

    A multiphysics analysis system for sodium-water reaction phenomena in a steam generator of sodium-cooled fast reactors was newly developed. The analysis system consists of the mechanistic numerical analysis codes, SERAPHIM, TACT, and RELAP5. The SERAPHIM code calculates the multicomponent multiphase flow and sodium-water chemical reaction caused by discharging of pressurized water vapor. Applicability of the SERAPHIM code was confirmed through the analyses of the experiment on water vapor discharging in liquid sodium. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The numerical models integrated into the TACT code were verified through some related experiments. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the tube rapidly heated by the reacting jet. The developed system enables evaluation of the wastage environment and the possibility of the failure propagation.

  19. Lead-Cooled Fast Reactor (LFR) Design: Safety, Neutronics, Thermal Hydraulics, Structural Mechanics, Fuel, Core, and Plant Design

    SciTech Connect

    Smith, C

    2010-02-22

    The idea of developing fast spectrum reactors with molten lead (or lead alloy) as a coolant is not a new one. Although initially considered in the West in the 1950s, such technology was not pursued to completion because of anticipated difficulties associated with the corrosive nature of these coolant materials. However, in the Soviet Union, such technology was actively pursued during the same time frame (1950s through the 1980s) for the specialized role of submarine propulsion. More recently, there has been a renewal of interest in the West for such technology, both for critical systems as well as for Accelerator Driven Subcritical (ADS) systems. Meanwhile, interest in the former Soviet Union, primarily Russia, has remained strong and has expanded well beyond the original limited mission of submarine propulsion. This section reviews the past and current status of LFR development.

  20. Fast cooling following a Late Triassic metamorphic and magmatic pulse: implications for the tectonic evolution of the Korean collision belt

    NASA Astrophysics Data System (ADS)

    de Jong, Koen; Han, Seokyoung; Ruffet, Gilles

    2015-11-01

    We discuss the evolution of Korea in the context of a relatively short-lived, tectonically induced, magmatic and metamorphic pulse that affected large portions of the crust of the peninsula's southern part during the Late Triassic. Recent 40Ar/39Ar single grain laser step-heating dates imply a prolonged metamorphic recrystallization between 243 and 220 Ma, which occurred in distinct phases that were not coeval throughout the peninsula. We obtained identical plateau ages between 231.4 ± 0.8 and 228.9 ± 0.8 Ma (1σ; 85-95% 39Ar release) on single grains of detrital muscovite from Jurassic sandstones (Gimpo Group). A literature review shows that the ages of detrital muscovites are identical to: (1) concordant 40Ar/39Ar ages of biotite (228 Ma) and amphibole (230 Ma) in amphibolites of the Deokjeongri Gneiss Formation and the Weolhyeonri Complex, pointing to very rapid cooling of 100-150 °C/Ma, and (2) 231-229 Ma muscovite from the low-grade metamorphic mid-Paleozoic turbidites of the Taean Formation. The efficiency of cooling is further underlined by the near-coincidence of these 40Ar/39Ar ages with 243-229 Ma (average: 234.6 Ma) zircon U-Pb ages in the Gyeonggi Massif and the Hongseong belt, in the literature. It is argued that the Late Triassic magmatic and metamorphic pulse is superimposed on an earlier tectono-metamorphic event, possibly related to collision, indicated by: (1) ~ 243-237 Ma muscovite ages, or age components in age spectra, and (2) two generations of folds and associated tectonic foliations truncated by ~ 229.5-Ma-old syenites and earlier mafic dykes. The Late Triassic thermal pulse could have been the result of post-collisional delamination of the lower crust and uppermost mantle, and/or oceanic slab break-off, which is also suggested by almost coeval, widespread mantle-sourced Mg-rich potassic magmatism. Continuing ductile deformation is shown by mylonitization of Late Triassic magmatic rocks; an ~ 220 Ma muscovite age may be related to this.

  1. The Relationship Between Microstructural Evolution and Mechanical Properties of Heavy Plate of Low-Mn Steel During Ultra Fast Cooling

    NASA Astrophysics Data System (ADS)

    Wang, Bin; Wang, Zhao-dong; Wang, Bing-xing; Wang, Guo-dong; Misra, R. D. K.

    2015-07-01

    We describe here the electron microscopy and mechanical property studies that were conducted in an industrially processed 20- and 40-mm C-Mn thick plates that involved a new approach of ultrafast cooling (UFC) together with significant reduction in Mn-content of the steel by ~0.3 to 0.5 pct, in relation to the conventional C-Mn steels, with the aim of cost-effectiveness. The study demonstrated that nanoscale cementite precipitation occurred during austenite transformation in the matrix of heavy plate during UFC, providing significant precipitation strengthening. With decrease in UFC stop temperature and consequent increase in the degree of undercooling, there was a transition in the morphology of cementite from lamellar to irregular-shaped nanoscale particles in the 20 mm heavy plate. With the increase in plate thickness, nanoscale cementite precipitated in bainitic lath at the surface of 40 mm heavy plate, which significantly increased the strength and decreased the elongation. Simultaneously, microstructural evolution in hot-rolled sheets was studied via simulation experiments using laboratory rolling mill to define the limits of microstructural evolution that can obtained in the UFC process and develop an understanding of the evolved microstructure in terms of process parameters.

  2. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, Progress Report for Work Through September 2002, 4th Quarterly Report

    SciTech Connect

    Mac Donald, Philip Elsworth

    2002-09-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If no additional moderator is added to the fuel rod lattice, it is possible to attain fast neutron energy spectrum conditions in a supercritical water-cooled reactor (SCWR). This type of core can make use of either fertile or fertile-free fuel and retain a hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity. One can also add moderation and design a thermal spectrum SCWR. The Generation IV Roadmap effort has identified the thermal spectrum SCWR (followed by the fast spectrum SCWR) as one of the advanced concepts that should be developed for future use. Therefore, the work in this NERI project is addressing both types of SCWRs.

  3. IODP Expedition 345: Structural characteristics of fast spread lower ocean crust, implications for growth and cooling of ocean crust

    NASA Astrophysics Data System (ADS)

    John, B. E.; Ceuleneer, G.; Cheadle, M. J.; Harigane, Y.

    2013-12-01

    IODP Expedition 345 to the Hess Deep Rift sampled ~1 Ma, fast-spread East Pacific Rise gabbroic crust exposed as a dismembered, lower crustal section. Sixteen holes were drilled at Site U1415, centered on a sub-horizontal, 200-m wide E-W-trending bench between 4675 and 4850 mbsl. The bench was formed as a rotational slide within a 1km high slump along the southern wall of the intra-rift ridge. Primitive olivine gabbro and troctolite (Mg# 76-89) were sampled in four discrete, 30 to ≥ 65 m sized blocks formed by the mass wasting that dominates the southwestern slope of the ridge. Igneous fabric orientations (both layering and foliation) in the blocks vary from sub-vertical to gently dipping, suggesting some of the blocks have rotated at least 90°. Magmatic fabrics including spectacular modal and/or grain size layering are prevalent in >50% of the recovered core. Magmatic foliation in all blocks is defined by plagioclase crystal shape, but may also be defined by olivine and, to a lesser extent, orthopyroxene and clinopyroxene when the crystals have suitable habits. In all cases, this foliation is controlled by both the preferred orientation and shape anisotropy of the crystals. Fabric intensity varies from moderate to strong in the block with simple modal layering, weak to absent in the two blocks of troctolite, and largely absent in the block with heterogeneous textures and/or diffuse banding. Intrinsic to the layering and banding is the common development of dendritic and/or skeletal olivine textures (grain size up to 3 cm). The preservation of these delicate olivine grains showing only limited subgrain formation, and no kinking precludes significant low melt fraction (<20%) crystal plastic flow of the cumulates. This observation prohibits ocean crust formation models that require homogeneous deformation/flow at low melt fractions. Down-temperature sub-solidus crystal plastic deformation and/or shear zones are virtually absent from the recovered core. Significant

  4. A CANDU-Based Fast Irradiation Reactor

    SciTech Connect

    Shatilla, Youssef

    2006-07-01

    A new steady-state fast neutron reactor is needed to satisfy the testing needs of Generation IV reactors, the Space Propulsion Program, and the Advanced Fuel Cycle Initiative. This paper presents a new concept for a CANDU-based fast irradiation reactor that is horizontal in orientation, with individual pressure tubes running the entire length of the scattering-medium tank (Calandria) filled with Lead-Bismuth-Eutectic (LBE). This approach for a test reactor will provide more flexibility in refueling, sample removal, and ability to completely re-configure the core to meet different users' requirements. Full core neutronic analysis of several fuel/coolant/geometry combinations showed a small hexagonal, LBE-cooled, U-Pu-10Zr fuel, with a core power of 100 MW{sub th} produced a fast flux (>0.1 MeV) of 1.5 x 10{sup 15} n/cm{sup 2} sec averaged over the whole length of six irradiation channels with a total testing volume of more than 77 liters. In-core breeding allowed the Pu-239 enrichment to be 15.3% which should result in core continuous operation for 180 effective full power days. Other coolants investigated included high pressure water steam and helium. An innovative shutdown/control system which consisted of the six outermost fuel channels was proven to be effective in shutting the core down when flooded with boric acid as a neutron absorber. The new shutdown/control system has the advantage of causing the minimum perturbation of the axial flux shape when the control channels are partially flooded with boric acid. This is because the acid is injected homogeneously along the control channel in contrast to regular control rods that are injected partially causing an axial perturbation in the core flux which in turn reduces safety analysis margins. The new shutdown/control system is not required to penetrate the core in a direction vertical to the fuel channels which allowed the freedom of changing core pitch as deemed necessary. A preliminary thermal hydraulic analysis

  5. The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input

    SciTech Connect

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal,; Monado, Fiber; Sekimoto, Hiroshi

    2012-06-06

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of i{sup th} region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.

  6. The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input

    NASA Astrophysics Data System (ADS)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Monado, Fiber; Sekimoto, Hiroshi

    2012-06-01

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of ith region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.

  7. Vitrification by Ultra-fast Cooling at a Low Concentration of Cryoprotectants in a Quartz Microcapillary: A Study Using Murine Embryonic Stem Cells

    PubMed Central

    He, Xiaoming; Park, Eric Y.H.; Fowler, Alex; Yarmush, Martin L.; Toner, Mehmet

    2009-01-01

    Conventional cryopreservation protocols for slow-freezing or vitrification involve cell injury due to ice formation/cell dehydration or toxicity of high cryoprotectant (CPA) concentrations, respectively. In this study, we developed a novel cryopreservation technique to achieve ultra-fast cooling rates using a quartz microcapillary (QMC). The QMC enabled vitrification of murine embryonic stem (ES) cells using an intracellular cryoprotectant concentration in the range used for slowing freezing (1–2 M). The cryoprotectants used included 2 M 1,2-propanediol (PROH, cell membrane permeable) and 0.5 M extracellular trehalose (cell membrane impermeable). More than 70% of the murine ES cells post-vitrification attached with respect to non-frozen control cells, and the proliferation rates of the two groups were similar. Preservation of undifferentiated properties of the pluripotent murine ES cells post vitrification cryopreservation was verified using three different types of assays: the expression of transcription factor Oct-4, the presentation of the membrane surface glycoprotein SSEA-1, and the elevated expression of the intracellular enzyme alkaline phosphatase. These results indicate that vitrification at a low concentration (2 M) of intracellular cryoprotectants is a viable and effective approach for the cryopreservation of murine embryonic stem cells. PMID:18462712

  8. Vitrification by ultra-fast cooling at a low concentration of cryoprotectants in a quartz micro-capillary: a study using murine embryonic stem cells.

    PubMed

    He, Xiaoming; Park, Eric Y H; Fowler, Alex; Yarmush, Martin L; Toner, Mehmet

    2008-06-01

    Conventional cryopreservation protocols for slow-freezing or vitrification involve cell injury due to ice formation/cell dehydration or toxicity of high cryoprotectant (CPA) concentrations, respectively. In this study, we developed a novel cryopreservation technique to achieve ultra-fast cooling rates using a quartz micro-capillary (QMC). The QMC enabled vitrification of murine embryonic stem (ES) cells using an intracellular cryoprotectant concentration in the range used for slowing freezing (1-2M). The cryoprotectants used included 2M 1,2-propanediol (PROH, cell membrane permeable) and 0.5M extracellular trehalose (cell membrane impermeable). More than 70% of the murine ES cells post-vitrification attached with respect to non-frozen control cells, and the proliferation rates of the two groups were similar. Preservation of undifferentiated properties of the pluripotent murine ES cells post-vitrification cryopreservation was verified using three different types of assays: the expression of transcription factor Oct-4, the presentation of the membrane surface glycoprotein SSEA-1, and the elevated expression of the intracellular enzyme alkaline phosphatase. These results indicate that vitrification at a low concentration (2M) of intracellular cryoprotectants is a viable and effective approach for the cryopreservation of murine embryonic stem cells. PMID:18462712

  9. Lead-cooled system design and challenges in the frame of Generation IV International Forum

    NASA Astrophysics Data System (ADS)

    Cinotti, Luciano; Smith, Craig F.; Sekimoto, Hiroshi; Mansani, Luigi; Reale, Marco; Sienicki, James J.

    2011-08-01

    The Generation IV International Forum (GIF) Technology Roadmap identified the Lead-cooled Fast Reactor (LFR) as a technology well suited for electricity generation, hydrogen production and actinide management in a closed fuel cycle. One of the most important features of the LFR is the fact that lead is a relatively inert coolant, a feature that conveys significant advantages in terms of safety, system simplification, and the consequent potential for economic performance. In 2004, the GIF LFR Provisional System Steering Committee was organized and began to develop the LFR System Research Plan. The committee selected two pool-type reactor concepts as candidates for international cooperation and joint development in the GIF framework: these are the Small Secure Transportable Autonomous Reactor (SSTAR); and the European Lead-cooled System (ELSY). The high boiling point (1745 °C) of lead has a beneficial impact to the safety of the system, whereas its high melting point (327.4 °C) requires new engineering strategies, especially for In-Service-Inspection and refuelling. Lead, especially at high temperatures, is also relatively corrosive towards structural materials. This necessitates that coolant purity and the level of dissolved oxygen be carefully controlled, in addition to the proper selection of structural materials. For the GIF LFR concepts, lead has been chosen as the coolant rather than Lead-Bismuth Eutectic primarily because of its greatly reduced generation of the alpha-emitting 210Po isotope formed in the coolant. This results in significantly reduced levels of radioactive contamination of the coolant while minimizing the effect of decay power in the coolant from such contaminants; an additional consideration is the desire to eliminate dependence on bismuth which might be a limited resource. This paper provides an overview of the historical development of the LFR, a summary of the advantages and challenges associated with heavy liquid metal coolants, and an

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

    NASA Astrophysics Data System (ADS)

    Ariani, Menik; Satya, Octavianus Cakra; Monado, Fiber; Su'ud, Zaki; Sekimoto, Hiroshi

    2016-03-01

    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).