WorldWideScience

Sample records for dust cooled reactors

  1. Gas cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1972-06-01

    Although most of the development work on fast breeder reactors has been devoted to the use of liquid metal cooling, interest has been expressed for a number of years in alternative breeder concepts using other coolants. One of a number of concepts in which interest has been retained is the Gas-Cooled Fast Reactor (GCFR). As presently envisioned, it would operate on the uranium-plutonium mixed oxide fuel cycle, similar to that used in the Liquid Metal Fast Breeder Reactor (LMFBR), and would use helium gas as the coolant.

  2. MEANS FOR COOLING REACTORS

    Science.gov (United States)

    Wheeler, J.A.

    1957-11-01

    A design of a reactor is presented in which the fuel elements may be immersed in a liquid coolant when desired without the necessity of removing them from the reactor structure. The fuel elements, containing the fissionable material are in plate form and are disposed within spaced slots in a moderator material, such as graphite to form the core. Adjacent the core is a tank containing the liquid coolant. The fuel elements are mounted in spaced relationship on a rotatable shaft which is located between the core and the tank so that by rotation of the shaft the fuel elements may be either inserted in the slots in the core to sustain a chain reaction or immersed in the coolant.

  3. Pebble Bed Reactor Dust Production Model

    Energy Technology Data Exchange (ETDEWEB)

    Abderrafi M. Ougouag; Joshua J. Cogliati

    2008-09-01

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

  4. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  5. Small Liquid Metal Cooled Reactor Safety Study

    Energy Technology Data Exchange (ETDEWEB)

    Minato, A; Ueda, N; Wade, D; Greenspan, E; Brown, N

    2005-11-02

    The Small Liquid Metal Cooled Reactor Safety Study documents results from activities conducted under Small Liquid Metal Fast Reactor Coordination Program (SLMFR-CP) Agreement, January 2004, between the Central Research Institute of the Electric Power Industry (CRIEPI) of Japan and the Lawrence Livermore National Laboratory (LLNL)[1]. Evaluations were completed on topics that are important to the safety of small sodium cooled and lead alloy cooled reactors. CRIEPI investigated approaches for evaluating postulated severe accidents using the CANIS computer code. The methods being developed are improvements on codes such as SAS 4A used in the US to analyze sodium cooled reactors and they depend on calibration using safety testing of metal fuel that has been completed in the TREAT facility. The 4S and the small lead cooled reactors in the US are being designed to preclude core disruption from all mechanistic scenarios, including selected unprotected transients. However, postulated core disruption is being evaluated to support the risk analysis. Argonne National Laboratory and the University of California Berkeley also supported LLNL with evaluation of cores with small positive void worth and core designs that would limit void worth. Assessments were also completed for lead cooled reactors in the following areas: (1) continuing operations with cladding failure, (2) large bubbles passing through the core and (3) recommendations concerning reflector control. The design approach used in the US emphasizes reducing the reactivity in the control mechanisms with core designs that have essentially no, or a very small, reactivity change over the core life. This leads to some positive void worth in the core that is not considered to be safety problem because of the inability to identify scenarios that would lead to voiding of lead. It is also believed that the void worth will not dominate the severe accident analysis. The approach used by 4S requires negative void worth throughout

  6. Study on water cooled high conversion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of study on advanced reactors for the future, conceptual design of high conversion water cooled reactors is being studied, aiming at the contribution to nuclear fuel cycle by the LWR technology, since the utilization of LWRs will extend over a long period of time . We are studying on the reactor core concepts for BWR and PWR reactor systems. As for BWR system, three types of reactor cores are investigating for three different design goals; long operation period, high conversion ratio and high applicability for the existing BWR system. In all the cases, we have obtained a fair prospect of a large core concept with a capacity of 1,000 MWe class having negative void reactivity coefficient. This study is a part of JAERI-JAPCO (Japan Atomic Power Company) cooperative studies. Various kinds of conceptual designs will be created until the end of FY 1999. The designs will be checked and reviewed at that time, then experimental studies on the realization of the concepts will start with further design works from FY 2000. (author)

  7. Lessons Learned From Gen I Carbon Dioxide Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    David E. Shropshire

    2004-04-01

    This paper provides a review of early gas cooled reactors including the Magnox reactors originating in the United Kingdom and the subsequent development of the Advanced Gas-cooled Reactors (AGR). These early gas cooled reactors shared a common coolant medium, namely carbon dioxide (CO2). A framework of information is provided about these early reactors and identifies unique problems/opportunities associated with use of CO2 as a coolant. Reactor designers successfully rose to these challenges. After years of successful use of the CO2 gas cooled reactors in Europe, the succeeding generation of reactors, called the High Temperature Gas Reactors (HTGR), were designed with Helium gas as the coolant. Again, in the 21st century, with the latest reactor designs under investigation in Generation IV, there is a revived interest in developing Gas Cooled Fast Reactors that use CO2 as the reactor coolant. This paper provides a historical perspective on the 52 CO2 reactors and the reactor programs that developed them. The Magnox and AGR design features and safety characteristics were reviewed, as well as the technologies associated with fuel storage, reprocessing, and disposal. Lessons-learned from these programs are noted to benefit the designs of future generations of gas cooled nuclear reactors.

  8. Liquid Metal Cooled Reactor for Space Power

    Science.gov (United States)

    Weitzberg, Abraham

    2003-01-01

    The conceptual design is for a liquid metal (LM) cooled nuclear reactor that would provide heat to a closed Brayton cycle (CBC) power conversion subsystem to provide electricity for electric propulsion thrusters and spacecraft power. The baseline power level is 100 kWe to the user. For long term power generation, UN pin fuel with Nb1Zr alloy cladding was selected. As part of the SP-100 Program this fuel demonstrated lifetime with greater than six atom percent burnup, at temperatures in the range of 1400-1500 K. The CBC subsystem was selected because of the performance and lifetime database from commercial and aircraft applications and from prior NASA and DOE space programs. The high efficiency of the CBC also allows the reactor to operate at relatively low power levels over its 15-year life, minimizing the long-term power density and temperature of the fuel. The scope of this paper is limited to only the nuclear components that provide heated helium-xenon gas to the CBC subsystem. The principal challenge for the LM reactor concept was to design the reactor core, shield and primary heat transport subsystems to meet mission requirements in a low mass configuration. The LM concept design approach was to assemble components from prior programs and, with minimum change, determine if the system met the objective of the study. All of the components are based on technologies having substantial data bases. Nuclear, thermalhydraulic, stress, and shielding analyses were performed using available computer codes. Neutronics issues included maintaining adequate operating and shutdown reactivities, even under accident conditions. Thermalhydraulic and stress analyses calculated fuel and material temperatures, coolant flows and temperatures, and thermal stresses in the fuel pins, components and structures. Using conservative design assumptions and practices, consistent with the detailed design work performed during the SP-100 Program, the mass of the reactor, shield, primary heat

  9. Smaller desert dust cooling effect estimated from analysis of dust size and abundance

    Science.gov (United States)

    Kok, Jasper F.; Ridley, David A.; Zhou, Qing; Miller, Ron L.; Zhao, Chun; Heald, Colette L.; Ward, Daniel S.; Albani, Samuel; Haustein, Karsten

    2017-03-01

    Desert dust aerosols affect Earth's global energy balance through direct interactions with radiation, and through indirect interactions with clouds and ecosystems. But the magnitudes of these effects are so uncertain that it remains unclear whether atmospheric dust has a net warming or cooling effect on global climate. Consequently, it is still uncertain whether large changes in atmospheric dust loading over the past century have slowed or accelerated anthropogenic climate change, or what the effects of potential future changes in dust loading will be. Here we present an analysis of the size and abundance of dust aerosols to constrain the direct radiative effect of dust. Using observational data on dust abundance, in situ measurements of dust optical properties and size distribution, and climate and atmospheric chemical transport model simulations of dust lifetime, we find that the dust found in the atmosphere is substantially coarser than represented in current global climate models. As coarse dust warms the climate, the global dust direct radiative effect is likely to be less cooling than the ~-0.4 W m-2 estimated by models in a current global aerosol model ensemble. Instead, we constrain the dust direct radiative effect to a range between -0.48 and +0.20 W m-2, which includes the possibility that dust causes a net warming of the planet.

  10. Initial prediction of dust production in pebble bed reactors

    Directory of Open Access Journals (Sweden)

    M. Rostamian

    2011-09-01

    Full Text Available This paper describes the computational simulation of contact zones between pebbles in a pebble bed reactor. In this type of reactor, the potential for graphite dust generation from frictional contact of graphite pebbles and the subsequent transport of dust and fission products can cause significant safety issues at very high temperatures around 900 °C in HTRs. The present simulation is an initial attempt to quantify the amount of nuclear grade graphite dust produced within a very high temperature reactor.

  11. Liquid metal cooled reactors for space power applications

    Science.gov (United States)

    Bailey, S.; Vaidyanathan, S.; Van Hoomissen, J.

    1985-01-01

    The technology basis for evaluation of liquid metal cooled space reactors is summarized. Requirements for space nuclear power which are relevant to selection of the reactor subsystem are then reviewed. The attributes of liquid metal cooled reactors are considered in relation to these requirements in the areas of liquid metal properties, neutron spectrum characteristics, and fuel form. Key features of typical reactor designs are illustrated. It is concluded that liquid metal cooled fast spectrum reactors provide a high confidence, flexible option for meeting requirements for SP-100 and beyond.

  12. Electrochemistry of Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

  13. COOL DUST IN THE OUTER RING OF NGC 1291

    Energy Technology Data Exchange (ETDEWEB)

    Hinz, J. L.; Engelbracht, C. W.; Skibba, R.; Montiel, E., E-mail: jhinz@as.arizona.edu [Steward Observatory, University of Arizona, 933 North Cherry Avenue, Tucson, AZ 85721 (United States); Crocker, A.; Calzetti, D. [Department of Astronomy, University of Massachusetts, Amherst, MA 01003 (United States); Donovan Meyer, J. [Physics and Astronomy Department, Stony Brook University, Stony Brook, NY 11794-3800 (United States); Sandstrom, K.; Walter, F.; Groves, B.; Meidt, S. E. [Max-Planck-Institut fur Astronomie, Konigstuhl 17, D-69117 Heidelberg (Germany); Johnson, B. D. [9 Institut d' Astrophysique de Paris, CNRS, Universite Pierre and Marie Curie, UMR 7095, 98bis bd Arago, F-75014 Paris (France); Hunt, L. [INAF-Osservatorio Astrofisico di Arcetri, Largo E. Fermi 5, I-50125 Fireze (Italy); Aniano, G.; Draine, B.; Murphy, E. J. [Department of Astrophysical Sciences, Princeton University, Princeton, NJ 08544 (United States); Armus, L. [Spitzer Science Center, California Institute of Technology, Mc 314-6, Pasadena, CA 91125 (United States); Dale, D. A. [Department of Physics and Astronomy, University of Wyoming, Laramie, WY 82071 (United States); Galametz, M.; Kennicutt, R. C. [Institute of Astronomy, University of Cambridge, Madingley Road, Cambridge CB3 0HA (United Kingdom); and others

    2012-09-01

    We examine Herschel Space Observatory images of one nearby prototypical outer ring galaxy, NGC 1291, and show that the ring becomes more prominent at wavelengths longer than 160 {mu}m. The mass of cool dust in the ring dominates the total dust mass of the galaxy, accounting for at least 70% of it. The temperature of the emitting dust in the ring (T = 19.5 {+-} 0.3 K) is cooler than that of the inner galaxy (T = 25.7 {+-} 0.7 K). We discuss several explanations for the difference in dust temperature, including age and density differences in the stellar populations of the ring versus the bulge.

  14. Cool Dust in the Outer Ring of NGC 1291

    CERN Document Server

    Hinz, J L; Skibba, R; Crocker, A; Meyer, J Donovan; Sandstrom, K; Walter, F; Montiel, E; Johnson, B D; Hunt, L; Aniano, G; Armus, L; Calzetti, D; Dale, D A; Draine, B; Galametz, M; Groves, B; Kennicutt, R C; Meidt, S E; Murphy, E J; Tabatabaei, F

    2012-01-01

    We examine Herschel Space Observatory images of one nearby prototypical outer ring galaxy, NGC 1291, and show that the ring becomes more prominent at wavelengths longer than 160um. The mass of cool dust in the ring dominates the total dust mass of the galaxy, accounting for at least 70% of it. The temperature of the emitting dust in the ring (T=19.5+/-0.3K) is cooler than that of the inner galaxy (T=25.7+/-0.7K). We discuss several explanations for the difference in dust temperature, including age and density differences in the stellar populations of the ring versus the bulge.

  15. Application of Hastelloy X in Gas-Cooled Reactor Systems

    DEFF Research Database (Denmark)

    Brinkman, C. R.; Rittenhouse, P. L.; Corwin, W.R.

    1976-01-01

    Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data...

  16. High Temperature Gas-Cooled Test Reactor Options Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

  17. Cooling Performance of Natural Circulation for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Suki; Chun, J. H.; Yum, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This paper deals with the core cooling performance by natural circulation during normal operation and a flow channel blockage event in an open tank-in-pool type research reactor. The cooling performance is predicted by using the RELAP5/ MOD3.3 code. The core decay heat is usually removed by natural circulation to the reactor pool water in open tank-in-pool type research reactors with the thermal power less than several megawatts. Therefore, these reactors have generally no active core cooling system against a loss of normal forced flow. In reactors with the thermal power less than around one megawatt, the reactor core can be cooled down by natural circulation even during normal full power operation. The cooling performance of natural circulation in an open tank-in-pool type research reactor has been investigated during the normal natural circulation and a flow channel blockage event. It is found that the maximum powers without void generation at the hot channel are around 1.16 MW and 820 kW, respectively, for the normal natural circulation and the flow channel blockage event.

  18. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    Baldev Raj

    2009-06-01

    The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective of providing fast reactor electricity at an affordable and competitive price.

  19. Dust divertor for a tokamak fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tang, X Z [Los Alamos National Laboratory; Delzanno, G L [Los Alamos National Laboratory

    2009-01-01

    Micron-size tungsten particulates find equilibrium position in the magnetized plasma sheath in the normal direction of the divertor surface, but are convected poloidally and toroidally by the sonic-ion-flow drag parallel to the divertor surface. The natural circulation of dust particles in the magnetized plasma sheath can be used to set up a flowing dust shield that absorbs and exhausts most of the tokamak heat flux to the divertor. The size of the particulates and the choice of materials offer substantial room for optimization.

  20. International working group on gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-15

    The purpose of the meeting was to provide a forum for exchange of information on safety and licensing aspects for gas-cooled reactors in order to provide comprehensive review of the present status and of directions for future applications and development. Contributions were made concerning the operating experience of the Fort St. Vrain (FSV) HTGR Power Plant in the United States of America, the experimental power station Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany, and the CO/sub 2/-cooled reactors in the United Kingdom such as Hunterson B and Hinkley Point B. The experience gained at each of these reactors has proved the high safety potential of Gas-cooled Reactor Power Plants.

  1. Fuel development for gas-cooled fast reactors

    Science.gov (United States)

    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.

  2. High power density reactors based on direct cooled particle beds

    Science.gov (United States)

    Powell, J. R.; Horn, F. L.

    Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

  3. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  4. Capital cost: gas cooled fast reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    The results of an investment cost study for a 900 MW(e) GCFR central station power plant are presented. The capital cost estimate arrived at is based on 1976 prices and a conceptual design only, not a mature reactor design.

  5. Simple analysis of an External Vessel Cooling Thermosyphon for a Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jae Young; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Song, Sub Lee [Handong Global University, Pohang (Korea, Republic of)

    2015-05-15

    KALIMER has three different DHR systems: two non-safety grade systems and one safety grade system. The non-safety grade systems are an IRACS (Intermediate Reactor Auxiliary Cooling System) and a steam/feedwater system. The safety grade system is a PDRC (Passive Decay Heat Removal Circuit). In case of the foreign reactor designs, ABTR (Advanced Burner Test Reactor) has a DRACS (Direct Reactor Auxiliary Cooling System), a PFBR (Indian Prototype Fast Breeder Reactor) has an SGDHRS (Safety Grade Decay Heat Removal System), and an EFR (European Fast Reactor) has DRC (Direct Reactor Cooling). Those designs have advantage on relatively high decay heat removal capacity. However, larger vessel size due to subsidiary in-vessel structure and possible accident propagation to reactor induced by sodium fire. In this paper, an ex-vessel thermosyphon design was proposed for the removal of decay heat for an iSFR. The proposed ex-vessel thermosyphon was designed to remove decay heat in both transient cases and BDBA cases, such as vessel failure. Proper working fluid was selected based on thermodynamic properties and chemical stability. Mercury was chosen as the working fluid, and SUS 314 was used for the corresponding structure material. Possible chemical reactions and adverse effects from using the thermosyphon were inherently eliminated by the system layout. A model for a high-temperature thermosyphon and numerical algorithms were used for the analysis. As a result of the simulation, the thermosyphon design was optimized, and it showed sufficient DHR performance to maintain core integrity.

  6. Study of Natural Convection Passive Cooling System for Nuclear Reactors

    Science.gov (United States)

    Abdillah, Habibi; Saputra, Geby; Novitrian; Permana, Sidik

    2017-07-01

    Fukushima nuclear reactor accident occurred due to the reactor cooling pumps and followed by all emergencies cooling systems could not work. Therefore, the system which has a passive safety system that rely on natural laws such as natural convection passive cooling system. In natural convection, the cooling material can flow due to the different density of the material due to the temperature difference. To analyze such investigation, a simple apparatus was set up and explains the study of natural convection in a vertical closed-loop system. It was set up that, in the closed loop, there is a heater at the bottom which is representing heat source system from the reactor core and cooler at the top which is showing the cooling system performance in room temperature to make a temperature difference for convection process. The study aims to find some loop configurations and some natural convection performances that can produce an optimum flow of cooling process. The study was done and focused on experimental approach and simulation. The obtained results are showing and analyzing in temperature profile data and the speed of coolant flow at some point on the closed-loop system.

  7. Gas-cooled reactors: the importance of their development

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.

    1979-06-01

    The nearest term GCR is the steam-cycle HTGR, which can be used for both power and process steam production. Use of SC-HTGRs permits timely introduction of thorium fuel cycles and of high-thermal-efficiency reactors, decreasing the need for mined U/sub 3/O/sub 8/ before arrival of symbiotic fueling of fast-thermal reactor systems. The gas-turbine HTGR offers prospects of lower capital costs than other nuclear reactors, but it appears to require longer and more costly development than the SC-HTGR. Accelerated development of the GT-HTGR is needed to gain the advantages of timely introduction. The Gas-Cooled Fast Breeder Reactor (GCFR) offers the possibility of fast breeder reactors with lower capital costs and with higher breeding ratios from oxide fuels. The VHTR provides high-temperature heat for hydrogen production.

  8. Simulation of Reactor Transient and Design Criteria of Sodium-cooled Fast Reactors

    OpenAIRE

    Gottfridsson, Filip

    2010-01-01

    The need for energy is growing in the world and the market of nuclear power is now once more expanding. Some issues of the current light-water reactors can be solved by the next generation of nuclear power, Generation IV, where sodium-cooled reactors are one of the candidates. Phénix was a French prototype sodium-cooled reactor, which is seen as a success. Although it did encounter an earlier unexperienced phenomenon, A.U.R.N., in which a negative reactivity transient followed by an oscillati...

  9. Control rod drive for high temperature gas cooled reactor

    Institute of Scientific and Technical Information of China (English)

    DengJun-Xian; XuJi-Ming; 等

    1998-01-01

    This control rod drive is developed for HTR-10 high temperature gas cooled test reactor.The stepmotor is prefered to improve positioning of the control rod and the scram behavior.The preliminary test in 1600170 ambient temperature shows that the selected stepmotor and transmission system can meet the main operation function requirements of HTR-10.

  10. Advanced gas cooled nuclear reactor materials evaluation and development program

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed.

  11. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs). DATES...

  12. System Study: Reactor Core Isolation Cooling 1998–2012

    Energy Technology Data Exchange (ETDEWEB)

    T. E. Wierman

    2013-10-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trend was identified in the HPCI results. Statistically significant decreasing trends were identified for RCIC start-only and 8-hour trends.

  13. System Study: Reactor Core Isolation Cooling 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-12-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  14. System Study: Reactor Core Isolation Cooling 1998–2013

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-01-31

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  15. Preliminary Assessment of a Debris Bed Cooling Performance for Demonstration Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Chung Ho; Park, Chang Gyu; Song, Hoon; Kim, Young Gyun; Jeong, Hae Yong; Chang, Jin Wook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    In the case of the sodium-cooled fast reactor such as KALIMER-600, Hypothetical Core Disruptive Accident (HCDA) attributed from mass nuclear fuel melting is unlikely to occur due to defense in depth concepts to meet requirements of redundancy and diversity. Multiple faults such as loss of flow, loss of heat sink, or transient overpower without scram are to lead rising the power level until cladding failure as reactivity increasing. The fact that metallic fuel melts at a lower temperature than the cladding allows significant in-pin- fuel motion to occur prior to cladding failure. Also, the combination of Doppler and axial expansion feedback and negative feedback associated with the in-pin fuel relocation prevents the reactivity from reaching prompt critical. Finally, the resulting reactivity and power reductions help prevent fuel temperatures from rising more than the fuel melting temperature. It is more difficult to occur HCDA in a metallic fueled core because reactor power and heat removal capability is maintained in balance by inherent safety characteristics However, for the future design of sodium-cooled fast reactor, the evaluation of the safety performance and the determination of containment requirements may be worth considering due to the triple-fault accident sequences of extremely low probability of occurrence that leads to core melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will be required as a design requirement for the future design of sodium cooled fast reactor. Also, proof of the capacity of the debris bed cooling is an essential condition to solve the problem of in-vessel retention of the core debris. Accordingly, evaluation of a packed debris bed cooling performance with single phase flow for demonstration sodium-cooled fast reactor was carried out for proof of the in-vessel retention of the core debris

  16. Superheated Water-Cooled Small Modular Underwater Reactor Concept

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-12-01

    Full Text Available A novel fully passive small modular superheated water reactor (SWR for underwater deployment is designed to produce 160 MWe with steam at 500ºC to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF. The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required.

  17. A gas-cooled reactor surface power system

    Science.gov (United States)

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  18. Impact of dust cooling on direct collapse black hole formation

    CERN Document Server

    Latif, M A; Habouzit, M; Schleicher, D R G; Volonteri, M

    2015-01-01

    Observations of quasars at $z> 6$ suggest the presence of black holes with a few times $\\rm 10^9 ~M_{\\odot}$. Numerous models have been proposed to explain their existence including the direct collapse which provides massive seeds of $\\rm 10^5~M_{\\odot}$. The isothermal direct collapse requires a strong Lyman-Werner flux to quench $\\rm H_2$ formation in massive primordial halos. In this study, we explore the impact of trace amounts of metals and dust enrichment. We perform three dimensional cosmological simulations for two halos of $\\rm > 10^7~M_{\\odot}$ with $\\rm Z/Z_{\\odot}= 10^{-4}-10^{-6}$ illuminated by an intense Lyman Werner flux of $\\rm J_{21}=10^5$. Our results show that initially the collapse proceeds isothermally with $\\rm T \\sim 8000$ K but dust cooling becomes effective at densities of $\\rm 10^{8}-10^{12} ~cm^{-3}$ and brings the gas temperature down to a few 100-1000 K for $\\rm Z/Z_{\\odot} \\geq 10^{-6}$. No gravitationally bound clumps are found in $\\rm Z/Z_{\\odot} \\leq 10^{-5}$ cases by the end...

  19. Convective cooling in a pool-type research reactor

    Science.gov (United States)

    Sipaun, Susan; Usman, Shoaib

    2016-01-01

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

  20. Convective cooling in a pool-type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sipaun, Susan, E-mail: susan@nm.gov.my [Malaysian Nuclear Agency, Industrial Technology Division, Blok 29T, Bangi 43200, Selangor (Malaysia); Usman, Shoaib, E-mail: usmans@mst.edu [Missouri University of Science and Technology, Nuclear Engineering, 222 Fulton Hall 301 W.14th St., Rolla 64509 MO (United States)

    2016-01-22

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.

  1. REVIEW OF REACTOR SAFETY ANALYSES OF FAST AND LIQUID METAL COOLED REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Shaver, R. E.; Wittenbrock, N. G.

    1967-11-01

    Safety analysis reports on United States fast and liquid metal cooled reactors were reviewed to gain a better understanding of the safety philosophy applied to the design of these facilities. This information was compiled to help guide the design and safety analysis of the Fast Flux Test Facility. No attempt was made to draw conclusions concerning the relative merit of different approaches and philosophies used by different reactor design teams. The facilities reviewed were; Enrico Fermi Atomic Power Plant (FERMI) Hallam Nuclear Power Facility (HALLAM) Southwest Experimental Fast Oxide Reactor (SEFOR) Fast Reactor Test Facility (FARET) Experimental Breeder Reactor No. 1 (EBR-I) Experimental Breeder Reactor No. 2 (EBR-II) Fast Reactor Zero Power Experiment (ZPR - III). The information gathered from the safety analysis reports is tabulated under these headings: Control and Safety Systems; Reactor Protection Systems; Backup Systems; Containment or Confinement Systems; Inherent Reactivity Effects and Important Physics Parameters; Fuel and Fuel Handling; Accidents Considered and Chemical Problems; Site; Exhaust Ventilation System; and Waste Effluents.

  2. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition

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

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

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

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

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

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

  5. Specific power of liquid-metal-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dobranich, D.

    1987-10-01

    Calculations of the core specific power for conceptual space-based liquid-metal-cooled reactors, based on heat transfer considerations, are presented for three different fuel types: (1) pin-type fuel; (2) cermet fuel; and (3) thermionic fuel. The calculations are based on simple models and are intended to provide preliminary comparative results. The specific power is of interest because it is a measure of the core mass required to produce a given amount of power. Potential problems concerning zero-g critical heat flux and loss-of-coolant accidents are also discussed because these concerns may limit the core specific power. Insufficient experimental data exists to accurately determine the critical heat flux of liquid-metal-cooled reactors in space; however, preliminary calculations indicate that it may be a concern. Results also indicate that the specific power of the pin-type fuels can be increased significantly if the gap between the fuel and the clad is eliminated. Cermet reactors offer the highest specific power because of the excellent thermal conductivity of the core matrix material. However, it may not be possible to take fuel advantage of this characteristic when loss-of-coolant accidents are considered in the final core design. The specific power of the thermionic fuels is dependent mainly on the emitter temperature. The small diameter thermionic fuels have specific powers comparable to those of pin-type fuels. 11 refs., 12 figs, 2 tabs.

  6. Delayed gamma power measurement for sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R., E-mail: romain.coulon@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Normand, S., E-mail: stephane.normand@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Ban, G., E-mail: ban@lpccaen.in2p3.f [ENSICAEN, 6 Boulevard Marechal Juin, F-14050 Caen Cedex 4 (France); Barat, E.; Montagu, T.; Dautremer, T. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Brau, H.-P. [ICSM, Centre de Marcoule, BP 17171 F-30207 Bagnols sur Ceze (France); Dumarcher, V. [AREVA NP, SET, F-84500 Bollene (France); Michel, M.; Barbot, L.; Domenech, T.; Boudergui, K.; Bourbotte, J.-M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Jousset, P. [CEA, LIST, Departement des Capteurs, du Signal et de l' Information, F-91191 Gif-sur-Yvette (France); Barouch, G.; Ravaux, S.; Carrel, F. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Saurel, N. [CEA, DAM, Laboratoire Mesure de Dechets et Expertise, F-21120 Is-sur-Tille (France); Frelin-Labalme, A.-M.; Hamrita, H. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France)

    2011-01-15

    Graphical abstract: Display Omitted Research highlights: {sup 20}F and {sup 23}Ne tagging agents are produced by fast neutron flux. {sup 20}F signal has been measured at the SFR Phenix prototype. A random error of only 3% for an integration time of 2 s could be achieved. {sup 20}F and {sup 23}Ne power measurement has a reduced temperature influence. Burn-up impact could be limited by simultaneous {sup 20}F and {sup 23}Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,{alpha}) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

  7. A Review of Dangerous Dust in Fusion Reactors: from Its Creation to Its Resuspension in Case of LOCA and LOVA

    Directory of Open Access Journals (Sweden)

    Andrea Malizia

    2016-07-01

    Full Text Available The choice of materials for the future nuclear fusion reactors is a crucial issue. In the fusion reactors, the combination of very high temperatures, high radiation levels, intense production of transmuting elements and high thermomechanical loads requires very high-performance materials. Erosion of PFCs (Plasma Facing Components determines their lifetime and generates a source of impurities (i.e., in-vessel tritium and dust inventories, which cool down and dilute the plasma. The resuspension of dust could be a consequences of LOss of Coolant Accidents (LOCA and LOss of Vacuum Accidents (LOVA and it can be dangerous because of dust radioactivity, toxicity, and capable of causing an explosion. These characteristics can jeopardize the plant safety and pose a serious threat to the operators. The purpose of this work is to determine the experimental and numerical steeps to develop a numerical model to predict the dust resuspension consequences in case of accidents through a comparison between the experimental results taken from campaigns carried out with STARDUST-U and the numerical simulation developed with CFD codes. The authors in this work will analyze the candidate materials for the future nuclear plants and the consequences of the resuspension of its dust in case of accidents through the experience with STARDUST-U.

  8. COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

    Directory of Open Access Journals (Sweden)

    Phani Kumar Domalapally

    2015-04-01

    Full Text Available Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

  9. Shape optimization of a sodium cooled fast reactor

    Science.gov (United States)

    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.

  10. Research and Development Roadmaps for Liquid Metal Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-04-20

    The United States Department of Energy (DOE) commissioned the development of technology roadmaps for advanced (non-light water reactor) reactor concepts to help focus research and development funding over the next five years. The roadmaps show the research and development needed to support demonstration of an advanced (non-LWR) concept by the early 2030s, consistent with DOE’s Vision and Strategy for the Development and Deployment of Advanced Reactors. The intent is only to convey the technical steps that would be required to achieve such a goal; the means by which DOE will determine whether to invest in specific tasks will be treated separately. The starting point for the roadmaps is the Technical Readiness Assessment performed as part of an Advanced Test and Demonstration Reactor study released in 2016. The roadmaps were developed based upon a review of technical reports and vendor literature summarizing the technical maturity of each concept and the outstanding research and development needs. Critical path tasks for specific systems were highlighted on the basis of time and resources needed to complete the tasks and the importance of the system to the performance of the reactor concept. The roadmaps are generic, i.e. not specific to a particular vendor’s design but vendor design information may have been used as representative of the concept family. In the event that both near-term and more advanced versions of a concept are being developed, either a single roadmap with multiple branches or separate roadmaps for each version were developed. In each case, roadmaps point to a demonstration reactor (engineering or commercial) and show the activities that must be completed in parallel to support that demonstration in the 2030-2035 window. This report provides the roadmaps for two fast reactor concepts, the Sodium-cooled Fast Reactor (SFR) and the Lead-cooled Fast Reactor (LFR). The SFR technology is mature enough for commercial demonstration by the early 2030s

  11. Application of Hastelloy X in gas-cooled reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Brinkman, C.R.; Rittenhouse, P.L.; Corwin, W.R.; Strizak, J.P.; Lystrup, A.; DiStefano, J.R.

    1976-10-01

    Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data are reported. Properties of concern include tensile, creep, creep-rupture, fatigue, creep-fatigue interaction, subcritical crack growth, thermal stability, and the influence of helium environments with controlled amounts of impurities on these properties. In order to develop these properties in helium environments that are expected to be prototypic of HTGR operating conditions, it was necessary to construct special environmental test systems. Details of construction and operating parameters are described. Interim results from tests designed to determine the above properties are presented. To date a fairly extensive amount of information has been generated on this material at Oak Ridge National Laboratory and elsewhere concerning behavior in air, which is reviewed. However, only limited data are available from tests conducted in helium. Comparisons of the fatigue and subcritical growth behavior in air between Hastelloy X and a number of other structural alloys are given.

  12. MELCOR Model Development of High Temperature Gas-cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Changyong; Huh, Changwook [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    The High Temperature Gas-cooled Reactor is one of the major challenging issues on the development of licensing technology for HTGR. The safety evaluation tools of HTGR can be developed in two ways - development of new HTGR-specific codes or revision of existing codes. The KINS is considering using existing analytic tools to the extent feasible, with appropriate modifications for the intended purpose. The system-level MELCOR code is traditionally used for LWR safety analysis, which is capable of performing thermal-fluid and accident analysis, including fission-product transport and release. Recently, this code is being modified for the NGNP HTGR by the NRC. In this study, the MELCOR input model for HTGR with Reactor Cavity Cooling System (RCCS) was developed and the steady state performance was analyzed to evaluate the applicability in HTGR. HTGR model with design characteristics of GT-MHR was developed using MELCOR 2.1 code to validate the applicability of MELCOR code to HTGR. In addition, the steady state of GT-MHR was analyzed with the developed model. It was evaluated to predict well the design parameters of GT-MHR. The developed model can be used as the basis for accident analysis of HTGR with further update of packages such as Radio Nuclide (RN) package.

  13. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

    OpenAIRE

    Pengcheng Zhao; Kangli Shi; Shuzhou Li; Jingchao Feng; Hongli Chen

    2016-01-01

    Small modular reactor (SMR) has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR) is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100) is being developed by University of Science and Technology of China (USTC). In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kineti...

  14. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira

    2013-01-16

    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  15. Description of the magnox type of gas cooled reactor (MAGNOX)

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, S.E.; Nonboel, E

    1999-05-01

    The present report comprises a technical description of the MAGNOX type of reactor as it has been build in Great Britain. The Magnox reactor is gas cooled (CO{sub 2}) with graphite moderators. The fuels is natural uranium in metallic form, canned with a magnesium alloy called 'Magnox'. The Calder Hall Magnox plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other stations are given in tables with a summary of design data. Special design features are also shortly described. Where specific data for Calder Hall Magnox has not been available, corresponding data from other Magnox plants has been used. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 sub-project 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au)

  16. Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-03-01

    This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

  17. CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin [Univ. of Wisconsin, Madison, WI (United Texas A & M Univ., College Station, TX (United States); Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

    2014-07-14

    The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

  18. Steam-Reheat Option for Supercritical-Water-Cooled Reactors

    Science.gov (United States)

    Saltanov, Eugene

    SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. Main objectives of the development are to increase thermal efficiency of a Nuclear Power Plant (NPP) and to decrease capital and operational costs. The first objective can be achieved by introducing nuclear steam reheat inside a reactor and utilizing regenerative feedwater heaters. The second objective can be achieved by designing a steam cycle that closely matches that of the mature supercritical fossil-fuelled power plants. The feasibility of these objectives is discussed. As a part of this discussion, heat-transfer calculations have been performed and analyzed for SuperCritical-Water (SCW) and SuperHeated-Steam (SHS) channels of the proposed reactor concept. In the calculations a uniform and three non-uniform Axial Heat Flux Profiles (AHFPs) were considered for six different fuels (UO2, ThO 2, MOX, UC2, UC, and UN) and at average and maximum channel power. Bulk-fluid, sheath, and fuel centerline temperatures as well as the Heat Transfer Coefficient (HTC) profiles were obtained along the fuel-channel length. The HTC values are within a range of 4.7--20 kW/m2·K and 9.7--10 kW/m2·K for the SCW and SHS channels respectively. The main conclusion is that while all the mentioned fuels may be used for the SHS channel, only UC2, UC, or UN are suitable for a SCW channel, because their fuel centerline temperatures are at least 1000°C below melting point, while that of UO2, ThO2 , and MOX may reach melting point.

  19. State of the art of nuclear facilities with organic cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brede, O.; Nagel, S.; Ziegenbein, D.

    1984-06-01

    USA, Canadian, and USSR activities aimed at developing nuclear facilities with organic cooled reactors are summarized. The facilities OMRE, PNPF, WR-1, and ARBUS are described, discussing in particular the problems of the chemistry of organic coolants. Finally, problems of further development and prospects of the application of organic cooled reactors are briefly outlined.

  20. Natural Convection and Boiling for Cooling SRP Reactors During Loss of Circulation Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Buckner, M.R.

    2001-06-26

    This study investigated natural convection and boiling as a means of cooling SRP reactors in the event of a loss of circulation accident. These studies show that single phase natural convection cooling of SRP reactors in shutdown conditions with the present piping geometry is probably not feasible.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  2. RCCS Experiments and Validation for High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh; Cliff Davis; Goon C. Park

    2007-09-01

    A reactor cavity cooling system (RCCS), an air-cooled helical coil RCCS unit immersed in the water pool, was proposed to overcome the disadvantages of the weak cooling ability of air-cooled RCCS and the complex structure of water-cooled RCCS for the high temperature gas-cooled reactor (HTGR). An experimental apparatus was constructed to investigate the various heat transfer phenomena in the water pool type RCCS, such as the natural convection of air inside the cavity, radiation in the cavity, the natural convection of water in the water pool and the forced convection of air in the cooling pipe. The RCCS experimental results were compared with published correlations. The CFX code was validated using data from the air-cooled portion of the RCCS. The RELAP5 code was validated using measured temperatures from the reactor vessel and cavity walls.

  3. Comparison of fuel assemblies in lead cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-15

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

  4. Study of reactor plant disturbed cooling condition modes caused by the VVER reactor secondary circuit

    Directory of Open Access Journals (Sweden)

    V.I. Belozerov

    2016-12-01

    Based on the RELAP-5, TRAC, and TRACE software codes, reactor plant cooling condition malfunction modes caused by the VVER-1000 secondary circuit were simulated and investigated. Experimental data on the mode with the turbine-generator stop valve closing are presented. The obtained dependences made it possible to determine the maximum values of pressure and temperature in the circulation circuit as well as estimate the Minimum Critical Heat Flux Ratio (MCHFR. It has been found that, if any of the initial events occurs, safety systems are activated according to the set points; transient processes are stabilized in time; and the Critical Heat Flux (CHF limit is provided. Therefore, in the event of emergency associated with the considered modes, the reactor plant safety will be ensured.

  5. Computational Fluid Dynamics Analysis of Very High Temperature Gas-Cooled Reactor Cavity Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Frisani, Angelo; Hassan, Yassin A; Ugaz, Victor M

    2010-11-02

    The design of passive heat removal systems is one of the main concerns for the modular very high temperature gas-cooled reactors (VHTR) vessel cavity. The reactor cavity cooling system (RCCS) is a key heat removal system during normal and off-normal conditions. The design and validation of the RCCS is necessary to demonstrate that VHTRs can survive to the postulated accidents. The computational fluid dynamics (CFD) STAR-CCM+/V3.06.006 code was used for three-dimensional system modeling and analysis of the RCCS. A CFD model was developed to analyze heat exchange in the RCCS. The model incorporates a 180-deg section resembling the VHTR RCCS experimentally reproduced in a laboratory-scale test facility at Texas A&M University. All the key features of the experimental facility were taken into account during the numerical simulations. The objective of the present work was to benchmark CFD tools against experimental data addressing the behavior of the RCCS following accident conditions. Two cooling fluids (i.e., water and air) were considered to test the capability of maintaining the RCCS concrete walls' temperature below design limits. Different temperature profiles at the reactor pressure vessel (RPV) wall obtained from the experimental facility were used as boundary conditions in the numerical analyses to simulate VHTR transient evolution during accident scenarios. Mesh convergence was achieved with an intensive parametric study of the two different cooling configurations and selected boundary conditions. To test the effect of turbulence modeling on the RCCS heat exchange, predictions using several different turbulence models and near-wall treatments were evaluated and compared. The comparison among the different turbulence models analyzed showed satisfactory agreement for the temperature distribution inside the RCCS cavity medium and at the standpipes walls. For such a complicated geometry and flow conditions, the tested turbulence models demonstrated that the

  6. The development and the tests of the electrostatic probe for dust particle collection in thermonuclear reactors

    Science.gov (United States)

    Begrambekov, L. B.; Voityuk, A. N.; Zakharov, A. M.

    2016-09-01

    Formation of dust particles in thermonuclear reactors can greatly affect the plasma parameters and lead to accumulation of tritium. The rates of formation and deposition of dust need to be measured, and the parameters of formation of dust particles and clusters need to be studied. A model of a device for collection of fine conductive particles capable of removing them from the reactor chamber for future research is proposed in this paper. The dust collector's operation is based on a principle of applied electrostatic field. The model was tested in different operating conditions: in vacuum, at the atmospheric pressure in the atmosphere of air and dry nitrogen. The experiments were conducted with a stationary system and with the dust collector in motion relative to the dusty surface. It is shown that, during the probe moving relative to the surface, it can remove up to 95% of fine tungsten particles with sizes ranging from 1 to 10 μm.

  7. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Greenwood, Michael Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  9. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

    Directory of Open Access Journals (Sweden)

    Pengcheng Zhao

    2016-01-01

    Full Text Available Small modular reactor (SMR has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100 is being developed by University of Science and Technology of China (USTC. In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS transient simulation at beginning of the reactor cycle (BOC has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

  10. High temperature gas-cooled reactor: gas turbine application study

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project.

  11. Dust Production and Mass Loss in Cool Evolved Stars

    Science.gov (United States)

    Boyer, M. L.

    2013-01-01

    Following the red giant branch phase and the subsequent core He-burning phase, the low- to intermediate-mass stars (0.8star. I will briefly review the current status of models that include AGB mass loss and relate them to recent observations of AGB stars from the Surveying the Agents of Galaxy Evolution (SAGE) Spitzer surveys of the Small and Large Magellanic Clouds, including measures of the total dust input to the interstellar medium from AGB stars.

  12. High Flux Isotopes Reactor (HFIR) Cooling Towers Demolition Waste Management

    Energy Technology Data Exchange (ETDEWEB)

    Pudelek, R. E.; Gilbert, W. C.

    2002-02-26

    This paper describes the results of a joint initiative between Oak Ridge National Laboratory, operated by UT-Battelle, and Bechtel Jacobs Company, LLC (BJC) to characterize, package, transport, treat, and dispose of demolition waste from the High Flux Isotope Reactor (HFIR), Cooling Tower. The demolition and removal of waste from the site was the first critical step in the planned HFIR beryllium reflector replacement outage scheduled. The outage was scheduled to last a maximum of six months. Demolition and removal of the waste was critical because a new tower was to be constructed over the old concrete water basin. A detailed sampling and analysis plan was developed to characterize the hazardous and radiological constituents of the components of the Cooling Tower. Analyses were performed for Resource Conservation and Recovery Act (RCRA) heavy metals and semi-volatile constituents as defined by 40 CFR 261 and radiological parameters including gross alpha, gross beta, gross gamma, alpha-emitting isotopes and beta-emitting isotopes. Analysis of metals and semi-volatile constituents indicated no exceedances of regulatory limits. Analysis of radionuclides identified uranium and thorium and associated daughters. In addition 60Co, 99Tc, 226Rm, and 228Rm were identified. Most of the tower materials were determined to be low level radioactive waste. A small quantity was determined not to be radioactive, or could be decontaminated. The tower was dismantled October 2000 to January 2001 using a detailed step-by-step process to aid waste segregation and container loading. The volume of waste as packaged for treatment was approximately 1982 cubic meters (70,000 cubic feet). This volume was comprised of plastic ({approx}47%), wood ({approx}38%) and asbestos transite ({approx}14%). The remaining {approx}1% consisted of the fire protection piping (contaminated with lead-based paint) and incidental metal from conduit, nails and braces/supports, and sludge from the basin. The waste

  13. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    OpenAIRE

    Galvez, Cristhian

    2011-01-01

    The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for electricity generation or other applications requiring the availability of heat at elevated temperatures. A stage in the design evolution of this plant requires the analysis of the plant during a variety of potential transients to understand the primary and safety cooling system response. This study focuses on the performance of the pa...

  14. Calculation of Radioactivity and Dose Rate of Activated Corrosion Products in Water-Cooled Fusion Reactor

    Directory of Open Access Journals (Sweden)

    Jingyu Zhang

    2016-01-01

    Full Text Available In water-cooled reactor, the dominant radioactive source term under normal operation is activated corrosion products (ACPs, which have an important impact on reactor inspection and maintenance. A three-node transport model of ACPs was introduced into the new version of ACPs source term code CATE in this paper, which makes CATE capable of theoretically simulating the variation and the distribution of ACPs in a water-cooled reactor and suitable for more operating conditions. For code testing, MIT PWR coolant chemistry loop was simulated, and the calculation results from CATE are close to the experimental results from MIT, which means CATE is available and credible on ACPs analysis of water-cooled reactor. Then ACPs in the blanket cooling loop of water-cooled fusion reactor ITER under construction were analyzed using CATE and the results showed that the major contributors are the short-life nuclides, especially Mn-56. At last a point kernel integration code ARShield was coupled with CATE, and the dose rate around ITER blanket cooling loop was calculated. Results showed that after shutting down the reactor only for 8 days, the dose rate decreased nearly one order of magnitude, which was caused by the rapid decay of the short-life ACPs.

  15. Closed Fuel Cycle and Minor Actinide Multirecycling in a Gas-Cooled Fast Reactor

    NARCIS (Netherlands)

    Van Rooijen, W.F.G.; Kloosterman, J.L.

    2009-01-01

    The Generation IV International Forum has identified the Gas-Cooled Fast Reactor (GCFR) as one of the reactor concepts for future deployment. The GCFR targets sustainability, which is achieved by the use of a closed nuclear fuel cycle where only fission products are discharged to a repository; all H

  16. Closed Fuel Cycle and Minor Actinide Multirecycling in a Gas-Cooled Fast Reactor

    NARCIS (Netherlands)

    Van Rooijen, W.F.G.; Kloosterman, J.L.

    2009-01-01

    The Generation IV International Forum has identified the Gas-Cooled Fast Reactor (GCFR) as one of the reactor concepts for future deployment. The GCFR targets sustainability, which is achieved by the use of a closed nuclear fuel cycle where only fission products are discharged to a repository; all

  17. Closed Fuel Cycle and Minor Actinide Multirecycling in a Gas-Cooled Fast Reactor

    NARCIS (Netherlands)

    Van Rooijen, W.F.G.; Kloosterman, J.L.

    2009-01-01

    The Generation IV International Forum has identified the Gas-Cooled Fast Reactor (GCFR) as one of the reactor concepts for future deployment. The GCFR targets sustainability, which is achieved by the use of a closed nuclear fuel cycle where only fission products are discharged to a repository; all H

  18. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  19. Effects of Nuclear Energy on Sustainable Development and Energy Security: Sodium-Cooled Fast Reactor Case

    OpenAIRE

    Sungjoo Lee; Byungun Yoon; Juneseuk Shin

    2016-01-01

    We propose a stepwise method of selecting appropriate indicators to measure effects of a specific nuclear energy option on sustainable development and energy security, and also to compare an energy option with another. Focusing on the sodium-cooled fast reactor, one of the highlighted Generation IV reactors, we measure and compare its effects with the standard pressurized water reactor-based nuclear power, and then with coal power. Collecting 36 indicators, five experts select seven key indic...

  20. Challenges and Innovative Technologies On Fuel Handling Systems for Future Sodium-Cooled Fast Reactors

    OpenAIRE

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

    2011-01-01

    International audience; The reactor refuelling system provides the means of transporting, storing, and handling reactor core subassemblies. The system consists of the facilities and equipment needed to accomplish the scheduled refuelling operations. The choice of a FHS impacts directly on the general design of the reactor vessel (primary vessel, storage, and final cooling before going to reprocessing), its construction cost, and its availability factor. Fuel handling design must take into acc...

  1. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  2. Steady Thermal Field Simulation of Forced Air-cooled Column-type Air-core Reactor

    Institute of Scientific and Technical Information of China (English)

    DENG Qiu; LI Zhenbiao; YIN Xiaogen; YUAN Zhao

    2013-01-01

    Modeling the steady thermal field of the column-type air-core reactor,and further analyzing its distribution regularity,will help optimizing reactor design as well as improving its quality.The operation mechanism and inner insulation structure of a novel current limiting column-type air-core reactor is introduced in this paper.The finite element model of five encapsulation forced air-cooled column type air-core reactor is constructed using Fluent.Most importantly,this paper present a new method that,the steady thermal field of reactor working under forced air-cooled condition is simulated without arbitrarily defining the convection heat transfer coefficient for the initial condition; The result of the thermal field distribution shows that,the maximum steady temperature rise of forced air-cooled columntype air-core reactor happens approximately 5% to its top.The law of temperature distribution indicates:In the 1/3part of the reactor to its bottom,the temperature will rise rapidly to the increasing of height,yet the gradient rate is gradually decreasing; In the 5 % part of the reactor to its top,the temperature will drop rapidly to the increasing of height; In the part between,the temperature will rise slowly to the increasing of height.The conclusion draws that more thermal withstand capacity should be considered at the 5 % part of the reactor to its top to achieve optimal design solution.

  3. Conceptual design of a passive moderator cooling system for a pressure tube type natural circulation boiling water cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Pal, Eshita, E-mail: eshi.pal@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Nayak, Arun K.; Vijayan, Pallipattu K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2015-09-15

    Highlights: • Passive moderator cooling system is designed to cool moderator passively during SBO. • PMCS is a system of two natural circulation loops, coupled via a heat exchanger. • RELAP5 analyses show that PMCS maintains moderator within safe limits for 7 days. - Abstract: The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days.

  4. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors AGENCY... Cooling Systems for New Boiling-Water Reactors.'' This RG describes testing methods the NRC staff considers acceptable for demonstrating the operability of emergency core cooling systems (ECCSs) for boiling...

  5. Studies on advanced water-cooled reactors beyond generation Ⅲ for power generation

    Institute of Scientific and Technical Information of China (English)

    CHENG Xu

    2007-01-01

    China's ambitious nuclear power program motivates the country's nuclear community to develop advanced reactor concepts beyond generation Ⅲ to ensure a long-term, stable, and sustainable development of nuclear power. The paper discusses some main criteria for the selection of future water-cooled reactors by considering the specific Chinese situation. Based on the suggested selection criteria, two new types of water-cooled reactors are recommended for future Chinese nuclear power generation. The high conversion pressurized water reactor utilizes the present PWR technology to a large extent. With a conversion ratio of about 0.95, the fuel utilization is increased about 5 times. This significantly improves the sustainability of fuel resources. The supercritical water-cooled reactor has favorable features in economics,sustainability and technology availability. It is a logical extension of the generation Ⅲ PWR technology in China.The status of international R&D work is reviewed. A new supercritieal water-cooled reactor (SCWR) core structure (the mixed reactor core) and a new fuel assembly design (two-rows FA) are proposed. The preliminary analysis using a coupled neutron-physics/thermal-hydranlics method is carded out. It shows good feasibility for the new design proposal.

  6. Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

    Science.gov (United States)

    Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

    2014-08-01

    The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

  7. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS`s heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis.

  8. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    Science.gov (United States)

    Sulaiman, S. A.; Dominguez-Ontiveros, E. E.; Alhashimi, T.; Budd, J. L.; Matos, M. D.; Hassan, Y. A.

    2015-04-01

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A&M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  9. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Sulaiman, S. A., E-mail: shamsulamri@tamu.edu; Dominguez-Ontiveros, E. E., E-mail: elvisdom@tamu.edu; Alhashimi, T., E-mail: jbudd123@tamu.edu; Budd, J. L., E-mail: dubaiboy@tamu.edu; Matos, M. D., E-mail: mailgoeshere@gmail.com; Hassan, Y. A., E-mail: yhasssan@tamu.edu [Department of Nuclear Engineering, Texas A and M University, College Station, TX, 77843-3133 (United States)

    2015-04-29

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  10. Data on test results of vessel cooling system of high temperature engineering test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Saikusa, Akio [Secretariat of Nuclear Safety Commission, Tokyo (Japan); Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2003-02-01

    High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28, 1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is the first Reactor Cavity Cooling System (RCCS) applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it was confirmed that the VCS heat removal at 30 MW power operation was higher than 0.3 MW. This paper shows outline of the VCS and test results on the VCS performance. (author)

  11. A review of gas-cooled reactor concepts for SDI (Strategic Defense Initiative) applications

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, A.C.

    1989-08-01

    We have completed a review of multimegawatt gas-cooled reactor concepts proposed for SDI applications. Our study concluded that the principal reason for considering gas-cooled reactors for burst-mode operation was the potential for significant system mass savings over closed-cycle systems if open-cycle gas-cooled operation (effluent exhausted to space) is acceptable. The principal reason for considering gas-cooled reactors for steady-state operation is that they may represent a lower technology risk than other approaches. In the review, nine gas-cooled reactor concepts were compared to identify the most promising. For burst-mode operation, the NERVA (Nuclear Engine for Rocket Vehicle Application) derivative reactor concept emerged as a strong first choice since its performance exceeds the anticipated operational requirements and the technology has been demonstrated and is retrievable. Although the NERVA derivative concepts were determined to be the lead candidates for the Multimegawatt Steady-State (MMWSS) mode as well, their lead over the other candidates is not as great as for the burst mode. 90 refs., 2 figs., 10 tabs.

  12. Safety Analysis of Pb-208 Cooled 800 MWt Modified CANDLE Reactors

    Science.gov (United States)

    Su'ud, Zaki; Widiawati, Nina; Sekimoto, H.; Artoto, A.

    2017-01-01

    Safely analysis of 800MWt Pb-208 cooled fast reactors with natural Uranium as fuel cycle input employing axial-radial combined Modiified CANDLE burnup scheme has been performed. The analysis of unprotected loss of flow(ULOF) and unprotected rod run-out transient overpower (UTOP) are discussed. Some simulations for 800 MWt Pb-208 cooled fast reactors has been performed and the results show that the reactor can anticipate complete pumping failure inherently by reducing power through reactivity feedback and remove the rest of heat through natural circulations. Compared to the Pb-nat cooled long life Fast Reactors, Pb-208 cooled reactors have smaller Doppler but higher coolant density reactivity coefficient. In the UTOP accident case the analysis has been performed against external reactivity up to 0.003dk/k. And for ULOHS case it is assumed that the secondary cooling system has broken. During all accident the cladding temperature is the most critical. Especially for the case of UTOP accident. In addition the steam generator design has also consider excess power which may reach 50% extra during severe UTOP case..

  13. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

    2009-10-01

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: • High temperature gas reactor fuels behavior • High temperature materials qualification • Design methods development and validation • Hydrogen production technologies • Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  14. Preliminary Study of Lead-Oxide Cooled Fast Reactor with Natural Uranium as an Input Fuel with Reactor Shuffling Strategy

    Science.gov (United States)

    Mahmudah, Rida SN; Su’ud, Zaki

    2017-01-01

    A preliminary study of lead-oxide cooled fast reactor with natural uranium as an input fuel using reactor shuffling strategy has been conducted. In this study, reactor core is divided into four zone with the same volume, each zone use different uranium enrichment. The enrichment number is estimated so that in the end of reactor’s operation, we only need to add natural uranium as the fresh input fuel. This study used UN-PuN as the fuel and lead oxide as the coolant. Several parameter studies have been conducted to determine the most suitable input condition. It is confirmed in this study that with fuel : cladding : coolant ratio of 53 : 10 : 37, and uranium enrichment in the first to the fourth zone of 0%, 6.25%, 7.5% and 8%, respectively, the reactor can operate as long as 20 years of operation with terminal k-eff of 1.0004.

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

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

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

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

    Directory of Open Access Journals (Sweden)

    Igor Shamanin

    2015-01-01

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

  17. Effects of Nuclear Energy on Sustainable Development and Energy Security: Sodium-Cooled Fast Reactor Case

    Directory of Open Access Journals (Sweden)

    Sungjoo Lee

    2016-09-01

    Full Text Available We propose a stepwise method of selecting appropriate indicators to measure effects of a specific nuclear energy option on sustainable development and energy security, and also to compare an energy option with another. Focusing on the sodium-cooled fast reactor, one of the highlighted Generation IV reactors, we measure and compare its effects with the standard pressurized water reactor-based nuclear power, and then with coal power. Collecting 36 indicators, five experts select seven key indicators to meet data availability, nuclear energy relevancy, comparability among energy options, and fit with Korean energy policy objectives. The results show that sodium-cooled fast reactors is a better alternative than existing nuclear power as well as coal electricity generation across social, economic and environmental dimensions. Our method makes comparison between energy alternatives easier, thereby clarifying consequences of different energy policy decisions.

  18. Cryogenic Cooling System for 5 kA, 200 μH Class HTS DC Reactor

    Science.gov (United States)

    Park, Heecheol; Kim, Seokho; Kim, Kwangmin; Park, Minwon; Park, Taejun; Kim, A.-rong; Lee, Sangjin

    DC reactors, made by aluminum busbar, are used to stabilize the arc of an electric furnace. In the conventional arc furnace, the transport current is several tens of kilo-amperes and enormous resistive loss is generated. To reduce the resistive loss at the DC reactor, a HTS DC reactor can be considered. It can dramatically improve the electric efficiency as well as reduce the installation space. Similar with other superconducting devices, the HTS DC reactor requires current leads from a power source in room temperature to the HTS coil in cryogenic environment. The heat loss at the metal current leads can be minimized through optimization process considering the geometry and the transport current. However, the transport current of the HTS DC reactor for the arc furnace is much larger than most of HTS magnets and the enormous heat penetration through the current lead should be effectively removed to keep the temperature around 70∼77 K. Current leads are cooled down by circulation of liquid nitrogen from the cooling system with a stirling cryocooler. The operating temperature of HTS coil is 30∼40 K and circulation of gaseous helium is used to remove the heat generation at the HTS coil. Gaseous helium is transported through the cryogenic helium blower and a single stage GM cryocooler. This paper describes design and experimental results on the cooling system for current leads and the HTS coil of 5 kA, 200 μH class DC reactor as a prototype. The results are used to verify the design values of the cooling systems and it will be applied to the design of scale-up cooling system for 50 kA, 200 μH class DC reactor.

  19. Circulating and plateout activity program for gas-cooled reactors with arbitrary radioactive chains

    Energy Technology Data Exchange (ETDEWEB)

    Apperson, C.E. Jr.

    1978-03-01

    A time-dependent method for estimating the fuel body, circulating, plateout, and filter inventory of a high temperature gas-cooled reactor (HTGR) during normal operation is discussed. The primary coolant model accounts for the source, buildup, decay, and cleanup of isotopes that are gas borne inside the prestressed concrete reactor vessel (PCRV). This method has been implemented in the SUVIUS computer program that is described in detail.

  20. Thermal-hydraulic code selection for modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Komen, E.M.J.; Bogaard, J.P.A. van den

    1995-06-01

    In order to study the transient thermal-hydraulic system behaviour of modular high temperature gas-cooled reactors, the thermal-hydraulic computer codes RELAP5, MELCOR, THATCH, MORECA, and VSOP are considered at the Netherlands Energy Research Foundation ECN. This report presents the selection of the most appropriate codes. To cover the range of relevant accidents, a suite of three codes is recommended for analyses of HTR-M and MHTGR reactors. (orig.).

  1. Device for cooling the main vessel of a fast fission nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Debru, M.

    1984-10-16

    The annular space delimited by the main vessel and an internal shell is in communication with the zone of the reactor vessel, in which the cold primary liquid is located. The annular space delimited by the shell and by an internal shell is in communication with the lower part of the core via tubes. Thus, the cold primary liquid is injected into the space where it circulates from bottom to top, and flows into the space, where it circulates from top to bottom while at the same time cooling the main vessel. The invention applies, in particular, to fast fission nuclear reactors cooled by liquid sodium.

  2. Enhancing VHTR Passive Safety and Economy with Thermal Radiation Based Direct Reactor Auxiliary Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Hongbin Zhang; Ling Zou; Xiaodong Sun

    2012-06-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The decay heat first is transferred to the core barrel by conduction and radiation, and then to the reactor vessel by thermal radiation and convection; finally the decay heat is transferred to natural circulated air or water systems. RVACS can be characterized as a surface based decay heat removal system. The RVACS is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to volume) and decay heat removal capability (proportional to surface area). When the relative decay heat removal capability decreases, the peak fuel temperature increases, even close to the design limit. Annular core designs with inner graphite reflector can mitigate this effect; therefore can further increase the reactor power. Another way to increase the reactor power is to increase power density. However, the reactor power is also limited by the decay heat removal capability. Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environment side. For the reactor side, cooling pipes will be inserted into holes made in the outer or

  3. Summarized compatibility review of reactor materials for CO2-cooled graphite-moderated reactors

    Energy Technology Data Exchange (ETDEWEB)

    Seddon, B.J.

    1964-09-23

    This report, which is a revised edition of TRG-Report-267, summarises an internal document and collates information on the compatibility of a range of materials used in CO{sub 2}-cooled graphite-moderated reactors. Information is presented in the form of six tables based on compatibilities of materials with carbon dioxide, beryllium, Magnox, magnesium, uranium and compatibilities of pairs of other relevant materials.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  5. Heat pipe cooled reactors for multi-kilowatt space power supplies

    Energy Technology Data Exchange (ETDEWEB)

    Ranken, W.A.; Houts, M.G.

    1995-01-01

    Three nuclear reactor space power system designs are described that demonstrate how the use of high temperature heat pipes for reactor heat transport, combined with direct conversion of heat to electricity, can result in eliminating pumped heat transport loops for both primary reactor cooling and heat rejection. The result is a significant reduction in system complexity that leads to very low mass systems with high reliability, especially in the power range of 1 to 20 kWe. In addition to removing heat exchangers, electromagnetic pumps, and coolant expansion chambers, the heat pipe/direct conversion combination provides such capabilities as startup from the frozen state, automatic rejection of reactor decay heat in the event of emergency or accidental reactor shutdown, and the elimination of single point failures in the reactor cooling system. The power system designs described include a thermoelectric system that can produce 1 to 2 kWe, a bimodal modification of this system to increase its power level to 5 kWe and incorporate high temperature hydrogen propulsion capability, and a moderated thermionic reactor concept with 5 to 20 kWe power output that is based on beryllium modules that thermally couple cylindrical thermionic fuel elements (TFEs) to radiator heat pipes.

  6. Heat pipe cooled reactors for multi-kilowatt space power supplies

    Science.gov (United States)

    Ranken, W. A.; Houts, M. G.

    Three nuclear reactor space power system designs are described that demonstrate how the use of high temperature heat pipes for reactor heat transport, combined with direct conversion of heat to electricity, can result in eliminating pumped heat transport loops for both primary reactor cooling and heat rejection. The result is a significant reduction in system complexity that leads to very low mass systems with high reliability, especially in the power range of 1 to 20 kWe. In addition to removing heat exchangers, electromagnetic pumps, and coolant expansion chambers, the heat pipe/direct conversion combination provides such capabilities as startup from the frozen state, automatic rejection of reactor decay heat in the event of emergency or accidental reactor shutdown, and the elimination of single point failures in the reactor cooling system. The power system designs described include a thermoelectric system that can produce 1 to 2 kWe, a bimodal modification of this system to increase its power level to 5 kWe and incorporate high temperature hydrogen propulsion capability, and a moderated thermionic reactor concept with 5 to 20 kWe power output that is based on beryllium modules that thermally couple cylindrical thermionic fuel elements (TFE's) to radiator heat pipes.

  7. Gas Cooled, Natural Uranium, D20 Moderated Power Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dahlberg, R.C.; Beasley, E.G.; DeBoer, T.K.; Evans, T.C.; Molino, D.F.; Rothwell, W.S.; Slivka, W.R.

    1956-08-01

    The attractiveness of a helium cooled, heavy water moderated, natural uranium central station power plant has been investigated. A fuel element has been devised which allows the D20 to be kept at a low pressure while the exit gas temperature is high. A preliminary cost analysis indicates that, using currently available materials, competitive nuclear power in foreign countries is possible.

  8. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

    Science.gov (United States)

    2013-10-25

    ... COMMISSION Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors AGENCY... Systems for Light-Water-Cooled Nuclear Power Reactors,'' in which the NRC made editorial corrections and... analysis for liquid and gaseous radwaste system components for light water nuclear power...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-03-01

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

  10. Application of objective provision tree to development of standard review plan for sodium-cooled fast reactor nuclear design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moo-Hoon; Suh, Namduk; Choi, Yongwon; Shin, Andong [Korea Institute of Nuclear Safety, Daejon (Korea, Republic of)

    2016-06-15

    A systematic methodology was developed for the standard review plan for sodium-cooled fast reactor nuclear design. The process is first to develop an objective provision tree of sodium-cooled fast reactor for the reactivity control safety function. The provision tree is generally developed by designer to confirm whether the design satisfies the defense-in-depth concept. Then applicability of the current standard review plan of nuclear design for light water reactor to sodium-cooled fast reactor was evaluated and complemented by the developed objective provision tree.

  11. MR-6 type fuel elements cooling in natural convection conditions after the reactor shut down

    Energy Technology Data Exchange (ETDEWEB)

    Pytel, K.; Bykowski, W.; Moldysz, A. [Institute of Atomic Energy, Otwock Swierk (Poland)

    2002-07-01

    Natural cooling conditions of the nuclear fuel in the channel type reactor after its shut down are commonly determined with relatively high uncertainty. This is not only to he lack of adequate measurements of thermal parameters i.e. the residual power generation, the coolant flow and temperatures, but also due to indeterminate model of convection mechanism. The numerical simulation of natural convection in multitube fuel assembly in the fuel channel leads to various convection modes including evidently chaotic behaviour. To determine the real cooling conditions in the MARIA research reactor a series of experiments has been performed with fuel assembly equipped with a set of thermocouples. After some forced cooling period (the shortest was half an hour after the reactor shut down) the reactor was left with the only natural convection. Two completely different cooling modes have been observed. The MARIA core consists of series of individual fuel channel and so called bypasses, maintaining the hydraulic properties of the fuel channel, connected in parallel. Initially, the convection cells were established trough few so-called bypasses providing a very effective mode of cooling. In this mode the flow charts were identical to those existing in forced cooling mode. After certain period the system switched on the second cooling mode with natural circulation within the individual fuel cells. Higher temperatures and temperature fluctuations were characteristic for this mode approaching 30 deg in amplitude. In almost all the cases the system was switching few times between modes, but eventually remained in the second mode. The switching times were not regular and the process has a chaotic behaviour. (author)

  12. Outflow dynamics of dust-driven wind models and implications for cool envelopes of PNe

    CERN Document Server

    Verbena, J L; Wachter, A

    2011-01-01

    The density profiles of cool envelopes of young Planetary Nebulae (PNe) are reminiscent of the final AGB outflow history of the central star, so far as these have not yet been transformed by the hot wind and radiation of the central star. Obviously, the evolution of the mass loss rate of that dust-driven, cool wind of the former giant in its final AGB stages must have shaped these envelopes to some extent. Less clear is the impact of changes in the outflow velocity. Certainly, larger and fast changes would lead to significant complications in the reconstruction of the mass-loss history from a cool envelope's density profile. Here, we analyse the outflow velocity v_{\\rm exp} in a consistent set of over 50 carbon-rich, dust-driven and well "saturated" wind models, and how it depends on basic stellar parameters. We find a relation of the kind of v_{\\rm exp} \\propto (L/M)^{0.6}. By contrast to the vast changes of the mass-loss rate in the final outflow phase, this relation suggest only very modest variations in t...

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  14. Safe design of cooled tubular reactors for exothermic multiple reactions: Multiple-reaction networks

    NARCIS (Netherlands)

    Westerink, E.J.; Westerterp, K.R.

    1988-01-01

    The model of the pseudo-homogeneous, one-dimensional cooled tubular reactor is applied to a multiple-reaction network. It is demonstrated for a network which consists of two parallel and two consecutive reactions. Three criteria are developed to obtain an integral yield which does not deviate more t

  15. Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-09-01

    The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure.

  16. Safe design of cooled tubular reactors for exothermic multiple reactions: Multiple-reaction networks

    NARCIS (Netherlands)

    Westerink, E.J.; Westerterp, K.R.

    1988-01-01

    The model of the pseudo-homogeneous, one-dimensional cooled tubular reactor is applied to a multiple-reaction network. It is demonstrated for a network which consists of two parallel and two consecutive reactions. Three criteria are developed to obtain an integral yield which does not deviate more

  17. Safe design of cooled tubular reactors for exothermic, multiple reactions. Consecutive reactions

    NARCIS (Netherlands)

    Westerterp, K.R.; Overtoom, R.R.M.

    1985-01-01

    The model of the pseudo-homogeneous, one-dimensional, cooled tubular reactor is applied to two consecutive, irreversible first order reactions. A criterion is derived to obtain a desired integral yield. Based on this criterion three requirements are formulated, which enable us to choose the relevant

  18. Thermally safe operation of a cooled semi-batch reactor: slow liquid-liquid reactions

    NARCIS (Netherlands)

    Steensma, M.; Westerterp, K.R.

    1988-01-01

    Thermally safe operation of a semi-batch reactor (SBR) implies that conditions leading to strong accumulation of unreacted reactants must be avoided. All thermal responses of a SBR, in which a slow liquid-liquid reaction takes place, can be represented in a diagram with the kinetics, cooling capacit

  19. Zirconium carbide coating for corium experiments related to water-cooled and sodium-cooled reactors

    Science.gov (United States)

    Plevacova, K.; Journeau, C.; Piluso, P.; Zhdanov, V.; Baklanov, V.; Poirier, J.

    2011-07-01

    Since the TMI and Chernobyl accidents the risk of nuclear severe accident is intensively studied for existing and future reactors. In case of a core melt-down accident in a nuclear reactor, a complex melt, called corium, forms. To be able to perform experiments with prototypic corium materials at high temperature, a coating which resists to different corium melts related to Generation I and II Water Reactors and Generation IV sodium fast reactor was researched in our experimental platforms both in IAE NNC in Kazakhstan and in CEA in France. Zirconium carbide was selected as protective coating for graphite crucibles used in our induction furnaces: VCG-135 and VITI. The method of coating application, called reactive wetting, was developed. Zirconium carbide revealed to resist well to the (U x, Zr y)O 2-z water reactor corium. It has also the advantage not to bring new elements to this chemical system. The coating was then tested with sodium fast reactor corium melts containing steel or absorbers. Undesirable interactions were observed between the coating and these materials, leading to the carburization of the corium ingots. Concerning the resistance of the coating to oxide melts without ZrO 2, the zirconium carbide coating keeps its role of protective barrier with UO 2-Al 2O 3 below 2000 °C but does not resist to a UO 2-Eu 2O 3 mixture.

  20. Design guide for category II reactors light and heavy water cooled reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

    Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

    1978-05-01

    The Department of Energy (DOE), in the ERDA Manual, requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification operation, maintainance, and decommissioning of DOW-owned reactors be in accordance with generally uniform standards, guide and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category II reactor structure, components, and systems.

  1. Safe design of cooled tubular reactors for exothermic, multiple reactions; parallel reactions—I: Development of criteria

    NARCIS (Netherlands)

    Westerterp, K.R.; Ptasiński, K.J.

    1984-01-01

    Previously reported design criteria for cooled tubular reactors are based on the prevention of reactor temperature run away and were developed for single reactions only. In this paper it is argued that such criteri a should be based on the reactor selectivity, from which eventually a maximum

  2. A preliminary safety analysis for the prototype Gen IV Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Jae Ho; Choi, Chi Woong; Jeong, Tae Kyeong; Ahn, Sang June; Lee, Seung Won; Chang, Won Pyo; Kang, Seok Hun; Yoo, Jae Woon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the in-vessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

  3. Stability analysis of supercritical-pressure light water-cooled reactor in constant pressure operation

    Energy Technology Data Exchange (ETDEWEB)

    Suhwan, JI; Shirahama, H.; Koshizuka, S.; Oka, Y. [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.

    2001-07-01

    The purpose of this study is to evaluate the thermal-hydraulic and the thermal-nuclear coupled stabilities of a supercritical pressure light water-cooled reactor. A stability analysis code at supercritical pressure is developed. Using this code, stabilities of full and partial-power reactor operating at supercritical pressure are investigated by the frequency-domain analysis. Two types of SCRs are analyzed; a supercritical light water reactor (SCLWR) and a supercritical water-cooled fast reactor (SCFR). The same stability criteria as Boiling Water Reactor are applied. The thermal-hydraulic stability of SCLWR and SCFR satisfies the criteria with a reasonable orifice loss coefficient. The decay ratio of the thermal-nuclear coupled stability in SCFR is almost zero because of a small coolant density coefficient of the fast reactor. The evaluated decay ratio of the thermal-nuclear coupled stability is 3,41 {approx} 10{sup -V} at 100% power in SCFR and 0,028 at 100% power in SCLWR. The sensitivity is investigated. It is found that the thermal-hydraulic stability is sensitive to the mass flow rate strongly and the thermal-nuclear coupled stability to the coolant density coefficient. The bottom power peak distribution makes the thermal-nuclear stability worse and the thermal-nuclear stability better. (author)

  4. Application of hafnium hydride control rod to large sodium cooled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ikeda, Kazumi, E-mail: kazumi_ikeda@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Moriwaki, Hiroyuki, E-mail: hiroyuki_moriwaki@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Ohkubo, Yoshiyuki, E-mail: yoshiyuki_okubo@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Iwasaki, Tomohiko, E-mail: tomohiko.iwasaki@qse.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Tohoku University, Aoba, Aramaki, Aoba-ku, Sendai-shi, Miyagi-ken 980-8579 (Japan); Konashi, Kenji, E-mail: konashi@imr.tohoku.ac.jp [Institute for Materials Research, Tohoku University, Narita-cho, Oarai-machi, Higashi-Ibaraki-gun, Ibaraki-ken 311-1313 (Japan)

    2014-10-15

    Highlights: • Application of hafnium hydride control rod to large sodium cooled fast breeder reactor. • This paper treats application of an innovative hafnium hydride control rod to a large sodium cooled fast breeder reactor. • Hydrogen absorption triples the reactivity worth by neutron spectrum shift at H/Hf ratio of 1.3. • Lifetime of the control rod quadruples because produced daughters of hafnium isotopes are absorbers. • Nuclear and thermal hydraulic characteristics of the reactor are as good as or better than B-10 enriched boron carbide. - Abstract: This study treats the feasibility of long-lived hafnium hydride control rod in a large sodium-cooled fast breeder reactor by nuclear and thermal analyses. According to the nuclear calculations, it is found that hydrogen absorption of hafnium triples the reactivity by the neutron spectrum shift at the H/Hf ratio of 1.3, and a hafnium transmutation mechanism that produced daughters are absorbers quadruples the lifetime due to a low incineration rate of absorbing nuclides under irradiation. That is to say, the control rod can function well for a long time because an irradiation of 2400 EFPD reduces the reactivity by only 4%. The calculation also reveals that the hafnium hydride control rod can apply to the reactor in that nuclear and thermal characteristics become as good as or better than 80% B-10 enriched boron carbide. For example, the maximum linear heat rate becomes 3% lower. Owing to the better power distribution, the required flow rate decreases approximately by 1%. Consequently, it is concluded on desk analyses that the long lived hafnium hydride control rod is feasible in the large sodium-cooled fast breeder reactor.

  5. Experimental investigation of a new method for advanced fast reactor shutdown cooling

    Science.gov (United States)

    Pakholkov, V. V.; Kandaurov, A. A.; Potseluev, A. I.; Rogozhkin, S. A.; Sergeev, D. A.; Troitskaya, Yu. I.; Shepelev, S. F.

    2017-07-01

    We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.

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

    Science.gov (United States)

    Palmer, David J.

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

  7. Gas-cooled thorium reactor with fuel block of the unified design

    Directory of Open Access Journals (Sweden)

    I.V. Shamanin

    2015-11-01

    Analysis of information materials pertaining to the use of thorium as fuel element in rector facilities of the new generation and of its future potential was performed in the present study. Results of the first phase of neutronics studies of 3D model of high-temperatures gas-cooled reactor facility on the basis of unified design of the fuel block are presented. Calculation 3D model was developed using the software code of the MCU-5 series. Several optimal configurations of the reactor core were selected according to the results of comparison of neutronics characteristics of the examined options for the purpose of development of small-size modular nuclear power installations with power up to 60MW. Results of calculations of reactivity margin of the reactor, neutron flux distribution and power density profiles are presented for the selected options of reactor core configuration.

  8. Effect of ion radiative cooling on Jeans instability of partially ionized dusty plasma with dust charge fluctuation

    Science.gov (United States)

    Sharma, Prerana; Patidar, Archana

    2017-01-01

    In this paper, the effect of ion radiative cooling on the gravitational instability of dusty plasma is studied, incorporating dust charge fluctuation with dust-neutral, neutral-ion, and ion-neutral collisions. The basic equations are linearized using normal mode analysis to obtain a general dispersion relation. The general dispersion relation is analytically and numerically discussed to explain the role of ion radiative cooling in the structure formation through gravitational instability. The Jeans collapse criteria are found to be modified due to ion and electron radiative cooling, dust charge fluctuations, and collisions effects. It is determined from the analytical and numerical calculations that the support of radiative cooling of ions drives thermal fluctuations and gives instability to the system. The electron cooling effect remains dominating over ion cooling effect, and thus, it enhances the collapse more efficiently than ion cooling effect. Although the radiative cooling is slow, it may precede the collapse in molecular cloud, which further leads to the structure formation. The present work is relevance for the structure formation in the molecular cloud.

  9. A 50-100 kWe gas-cooled reactor for use on Mars.

    Energy Technology Data Exchange (ETDEWEB)

    Peters, Curtis D. (.)

    2006-04-01

    In the space exploration field there is a general consensus that nuclear reactor powered systems will be extremely desirable for future missions to the outer solar system. Solar systems suffer from the decreasing intensity of solar radiation and relatively low power density. Radioisotope Thermoelectric Generators are limited to generating a few kilowatts electric (kWe). Chemical systems are short-lived due to prodigious fuel use. A well designed 50-100 kWe nuclear reactor power system would provide sufficient power for a variety of long term missions. This thesis will present basic work done on a 50-100 kWe reactor power system that has a reasonable lifespan and would function in an extraterrestrial environment. The system will use a Gas-Cooled Reactor that is directly coupled to a Closed Brayton Cycle (GCR-CBC) power system. Also included will be some variations on the primary design and their effects on the characteristics of the primary design. This thesis also presents a variety of neutronics related calculations, an examination of the reactor's thermal characteristics, feasibility for use in an extraterrestrial environment, and the reactor's safety characteristics in several accident scenarios. While there has been past work for space reactors, the challenges introduced by thin atmospheres like those on Mars have rarely been considered.

  10. Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

    Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

  11. Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH{sub 2} and B{sub 4}C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor.

  12. Preliminary Reactor Head Bolt Design of Prototype Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Han, Insu; Koo, Gyeonghoi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    As structural requirements, the reactor head is designed to withstand all of the pressure, temperatures and forces which are likely to be imposed on it. The bolts that fasten the head to the vessel flange. Design of the reactor head bolts so as to withstand the loads applied should be designed. Currently, preliminary design of the PGSFR reactor bolts is progressed. So far, we have designed and evaluated example. The number and cross-sectional areas of bolts were determined using the procedure given in ASME BPVC Section III, Division 1, Appendix E. The purpose of this study is to conduct design the number and cross-sectional area of bolts attaching the PGSFR reactor head to the reactor vessel, using the ASME procedure. In this paper, preliminary bolt design for PGSFR was carried out according to the ASME procedure. Detailed calculations were carried out for bolt root diameter = 80 mm and number of bolts Nb = 45. It should be noted that the seating pressure recommended in the ASME code is only a suggested value, not mandatory appendix E. It does not guarantee a leak-tight joint. So these quantities are needed to carry out fatigue analysis of the bolts and to assure leak tightness of the joint during operation. For the future work, the fatigue and seismic analysis will be performed.

  13. SPIRE imaging of M82: cool dust in the wind and tidal streams

    CERN Document Server

    Roussel, H; Vigroux, L; Isaak, K G; Sauvage, M; Madden, S C; Auld, R; Baes, M; Barlow, M J; Bendo, G J; Bock, J J; Boselli, A; Bradford, M; Buat, V; Castro-Rodriguez, N; Chanial, P; Charlot, S; Ciesla, L; Clements, D L; Cooray, A; Cormier, D; Cortese, L; Davies, J I; Dwek, E; Eales, S A; Elbaz, D; Galametz, M; Galliano, F; Gear, W K; Glenn, J; Gomez, H L; Griffin, M; Hony, S; Levenson, L R; Lu, N; O'Halloran, B; Okumura, K; Oliver, S; Page, M J; Panuzzo, P; Papageorgiou, A; Parkin, T J; Perez-Fournon, I; Pohlen, M; Rangwala, N; Rigby, E E; Rykala, A; Sacchi, N; Schulz, B; Schirm, M R P; Smith, M W L; Spinoglio, L; Stevens, J A; Srinivasan, S; Symeonidis, M; Trichas, M; Vaccari, M; Wozniak, H; Wright, G S; Zeilinger, W W

    2010-01-01

    M82 is a unique representative of a whole class of galaxies, starbursts with superwinds, in the Very Nearby Galaxy Survey with Herschel. In addition, its interaction with the M81 group has stripped a significant portion of its interstellar medium from its disk. SPIRE maps now afford better characterization of the far-infrared emission from cool dust outside the disk, and sketch a far more complete picture of its mass distribution and energetics than previously possible. They show emission coincident in projection with the starburst wind and in a large halo, much more extended than the PAH band emission seen with Spitzer. Some complex substructures coincide with the brightest PAH filaments, and others with tidal streams seen in atomic hydrogen. We subtract the far-infrared emission of the starburst and underlying disk from the maps, and derive spatially-resolved far-infrared colors for the wind and halo. We interpret the results in terms of dust mass, dust temperature, and global physical conditions. In partic...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-01

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

  15. Optimization of Internal Cooling Fins for Metal Hydride Reactors

    Directory of Open Access Journals (Sweden)

    Vamsi Krishna Kukkapalli

    2016-06-01

    Full Text Available Metal hydride alloys are considered as a promising alternative to conventional hydrogen storage cylinders and mechanical hydrogen compressors. Compared to storing in a classic gas tank, metal hydride alloys can store hydrogen at nearly room pressure and use less volume to store the same amount of hydrogen. However, this hydrogen storage method necessitates an effective way to reject the heat released from the exothermic hydriding reaction. In this paper, a finned conductive insert is adopted to improve the heat transfer in the cylindrical reactor. The fins collect the heat that is volumetrically generated in LaNi5 metal hydride alloys and deliver it to the channel located in the center, through which a refrigerant flows. A multiple-physics modeling is performed to analyze the transient heat and mass transfer during the hydrogen absorption process. Fin design is made to identify the optimum shape of the finned insert for the best heat rejection. For the shape optimization, use of a predefined transient heat generation function is proposed. Simulations show that there exists an optimal length for the fin geometry.

  16. Impact of Dust Cooling on Direct-collapse Black Hole Formation

    Science.gov (United States)

    Latif, M. A.; Omukai, K.; Habouzit, M.; Schleicher, D. R. G.; Volonteri, M.

    2016-05-01

    Observations of quasars at z\\gt 6 suggest the presence of black holes with a few times {10}9\\quad {M}⊙ . Numerous models have been proposed to explain their existence, including a direct collapse, which provides massive seeds of {10}5\\quad {M}⊙ . The isothermal direct collapse requires a strong Lyman-Werner (LW) flux to quench {{{H}}}2 formation in massive primordial halos. In this study, we explore the impact of trace amounts of metals and dust enrichment. We perform three-dimensional cosmological simulations for two halos of \\gt {10}7\\quad {M}⊙ with Z/{Z}⊙ ={10}-4{--}{10}-6 illuminated by an intense LW flux of {J}21={10}5. Our results show that initially the collapse proceeds isothermally with T˜ 8000 K, but dust cooling becomes effective at densities of {10}8{--}{10}12 {{cm}}-3 and brings the gas temperature down to a few 100-1000 K for Z/{Z}⊙ ≥slant {10}-6. No gravitationally bound clumps are found in the Z/{Z}⊙ ≤slant {10}-5 cases by the end of our simulations, in contrast to the case with Z/{Z}⊙ ={10}-4. Large inflow rates of ≥slant 0.1\\quad {M}⊙ {{yr}}-1 are observed for Z/{Z}⊙ ≤slant {10}-5, similar to a zero-metallicity case, while for Z/{Z}⊙ ={10}-4 the inflow rate starts to decline earlier because of dust cooling and fragmentation. For given large inflow rates, a central star of ˜ {10}4\\quad {M}⊙ may form for Z/{Z}⊙ ≤slant {10}-5.

  17. Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Barsell, A.W.

    1980-05-01

    This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core.

  18. Core Design and Deployment Strategy of Heavy Water Cooled Sustainable Thorium Reactor

    Directory of Open Access Journals (Sweden)

    Naoyuki Takaki

    2012-08-01

    Full Text Available Our previous studies on water cooled thorium breeder reactor based on matured pressurized water reactor (PWR plant technology concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array and using heavy water as coolant is appropriate for achieving better breeding performance and higher burn-up simultaneously [1–6]. One optimum core that produces 3.5 GW thermal energy using Th-233U oxide fuel shows a breeding ratio of 1.07 and averaged burn-up of about 80 GWd/t with long cycle length of 1300 days. The moderator to fuel volume ratio is 0.6 and required enrichment of 233U for the fresh fuel is about 7%. The coolant reactivity coefficient is negative during all cycles despite it being a large scale breeder reactor. In order to introduce this sustainable thorium reactor, three-step deployment scenario, with intermediate transition phase between current light water reactor (LWR phase and future sustainer phase, is proposed. Both in transition phase and sustainer phase, almost the same core design can be applicable only by changing fissile materials mixed with thorium from plutonium to 233U with slight modification in the fuel assembly design. Assuming total capacity of 60 GWe in current LWR phase and reprocessing capacity of 800 ton/y with further extensions to 1600 ton/y, all LWRs will be replaced by heavy water cooled thorium reactors within about one century then thorium reactors will be kept operational owing to its potential to sustain fissile fuels while reprocessing all spent fuels until exhaustion of massive thorium resource.

  19. Water cooled breeder program summary report (LWBR (Light Water Breeder Reactor) development program)

    Energy Technology Data Exchange (ETDEWEB)

    1987-10-01

    The purpose of the Department of Energy Water Cooled Breeder Program was to demonstrate pratical breeding in a uranium-233/thorium fueled core while producing electrical energy in a commercial water reactor generating station. A demonstration Light Water Breeder Reactor (LWBR) was successfully operated for more than 29,000 effective full power hours in the Shippingport Atomic Power Station. The reactor operated with an availability factor of 76% and had a gross electrical output of 2,128,943,470 kilowatt hours. Following operation, the expended core was examined and no evidence of any fuel element defects was found. Nondestructive assay of 524 fuel rods determined that 1.39 percent more fissile fuel was present at the end of core life than at the beginning, proving that breeding had occurred. This demonstrates the existence of a vast source of electrical energy using plentiful domestic thorium potentially capable of supplying the entire national need for many centuries. To build on the successful design and operation of the Shippingport Breeder Core and to provide the technology to implement this concept, several reactor designs of large breeders and prebreeders were developed for commercial-sized plants of 900--1000 Mw(e) net. This report summarizes the Water Cooled Breeder Program from its inception in 1965 to its completion in 1987. Four hundred thirty-six technical reports are referenced which document the work conducted as part of this program. This work demonstrated that the Light Water Breeder Reactor is a viable alternative as a PWR replacement in the next generation of nuclear reactors. This transition would only require a minimum of change in design and fabrication of the reactor and operation of the plant.

  20. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Quiping [The Ohio State Univ., Columbus, OH (United States); Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Chtistensen, Richard [The Ohio State Univ., Columbus, OH (United States); Blue, Thomas [The Ohio State Univ., Columbus, OH (United States); Yoder, Graydon [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-08

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  1. Phytoplankton distribution in three thermally different but edaphically similar reactor cooling reservoirs

    Energy Technology Data Exchange (ETDEWEB)

    Wilde, E W

    1982-01-01

    Phytoplankton community structure and the physicochemical characteristics of three reactor cooling reservoirs in close proximity and of similar age and bottom type were studied during 1978. The three reservoirs differed in thermal alteration resulting from reactor cooling water as follows: (1) considerable heating with lake-wide temperatures >30/sup 0/C, even in winter; (2) a maximal 5/sup 0/C increase occurring in only one of three major arms of the reservoir; and (3) no thermal effluent received during the study period. Considerable spatial and temporal differences in water quality and phytoplankton community structure were observed; however, water temperature independent of other environmental factors (e.g., light and nutrients) was found to be a relatively unimportant variable for explaining phytoplankton periodicity.

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

    Directory of Open Access Journals (Sweden)

    Lap-Yan Cheng

    2009-01-01

    Full Text Available 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.

  3. An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

    2009-11-01

    This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

  4. Monte Carlo modeling of Lead-Cooled Fast Reactor in adiabatic equilibrium state

    Energy Technology Data Exchange (ETDEWEB)

    Stanisz, Przemysław, E-mail: pstanisz@agh.edu.pl; Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

    2016-05-15

    Graphical abstract: - Highlights: • We present the Monte Carlo modeling of the LFR in the adiabatic equilibrium state. • We assess the adiabatic equilibrium fuel composition using the MCB code. • We define the self-adjusting process of breeding gain by the control rod operation. • The designed LFR can work in the adiabatic cycle with zero fuel breeding. - Abstract: Nuclear power would appear to be the only energy source able to satisfy the global energy demand while also achieving a significant reduction of greenhouse gas emissions. Moreover, it can provide a stable and secure source of electricity, and plays an important role in many European countries. However, nuclear power generation from its birth has been doomed by the legacy of radioactive nuclear waste. In addition, the looming decrease in the available resources of fissile U235 may influence the future sustainability of nuclear energy. The integrated solution to both problems is not trivial, and postulates the introduction of a closed-fuel cycle strategy based on breeder reactors. The perfect choice of a novel reactor system fulfilling both requirements is the Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state. In such a state, the reactor converts depleted or natural uranium into plutonium while consuming any self-generated minor actinides and transferring only fission products as waste. We present the preliminary design of a Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state with the Monte Carlo Continuous Energy Burnup Code – MCB. As a reference reactor model we apply the core design developed initially under the framework of the European Lead-cooled SYstem (ELSY) project and refined in the follow-up Lead-cooled European Advanced DEmonstration Reactor (LEADER) project. The major objective of the study is to show to what extent the constraints of the adiabatic cycle are maintained and to indicate the phase space for further improvements. The analysis

  5. A modular gas-cooled cermet reactor system for planetary base power

    Science.gov (United States)

    Jahshan, Salim N.; Borkowski, Jeffrey A.

    1993-01-01

    Fission nuclear power is foreseen as the source for electricity in planetary colonization and exploration. A six module gas-cooled, cermet-fueled reactor is proposed that can meet the design objectives. The highly enriched core is compact and can operate at high temperature for a long life. The helium coolant powers six modular Brayton cycles that compare favorably with the SP-100-based Brayton cycle.

  6. Thermally safe operation of a cooled semi-batch reactor: slow liquid-liquid reactions

    OpenAIRE

    Steensma, M.; Westerterp, K R

    1988-01-01

    Thermally safe operation of a semi-batch reactor (SBR) implies that conditions leading to strong accumulation of unreacted reactants must be avoided. All thermal responses of a SBR, in which a slow liquid-liquid reaction takes place, can be represented in a diagram with the kinetics, cooling capacity and potential temperature rise as the keyfactors. Slow reactions taking place in the dispersed phase were found to be more prone to accumulation than reactions in the continuous phase. An overhea...

  7. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  8. Analysis of Precooling Injection Transient of Steam Generator for High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Yan Wang

    2017-01-01

    Full Text Available After a postulated design basis accident leads high temperature gas cooled reactor to emergency shutdown, steam generator still remains with high temperature level and needs to be cooled down by a precooling before reactor restarts with clearing of fault. For the large difference of coolant temperature between inlet and outlet of steam generator in normal operation, the temperature distribution on the components of steam generator is very complicated. Therefore, the temperature descending rate of the components in steam generator needs to be limited to avoid the potential damage during the precooling stage. In this paper, a pebble-bed high temperature gas cooled reactor is modeled by thermal-hydraulic system analysis code and several postulated precooling injection transients are simulated and compared to evaluate their effects, which will provide support for the precooling design. The analysis results show that enough precooling injection is necessary to satisfy the precooling requirements, and larger mass flow rate of precooling water injection will accelerate the precooling process. The temperature decrease of steam generator is related to the precooling injection scenarios, and the maximal mass flow rate of the precooling injection should be limited to avoid the excessively quick temperature change of the structures in steam generator.

  9. Simulation and control of water-gas shift packed bed reactor with inter-stage cooling

    Science.gov (United States)

    Saw, S. Z.; Nandong, J.

    2016-03-01

    Water-Gas Shift Reaction (WGSR) has become one of the well-known pathways for H2 production in industries. The issue with WGSR is that it is kinetically favored at high temperatures but thermodynamically favored at low temperatures, thus requiring careful consideration in the control design in order to ensure that the temperature used does not deactivate the catalyst. This paper studies the effect of a reactor arrangement with an inter-stage cooling implemented in the packed bed reactor to look at its effect on outlet temperature. A mathematical model is developed based on one-dimensional heat and mass transfers which incorporate the intra-particle effects. It is shown that the placement of the inter-stage cooling and the outlet temperature exiting the inter-stage cooling have strong influence on the reaction conversion. Several control strategies are explored for the process. It is shown that a feedback- feedforward control strategy using Multi-scale Control (MSC) is effective to regulate the reactor temperature profile which is critical to maintaining the catalysts activity.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-02-09

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

  12. Fluoride Salt-Cooled High-Temperature Reactor Technology Development and Demonstration Roadmap

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Mays, Gary T [ORNL; Pointer, William David [ORNL; Robb, Kevin R [ORNL; Yoder Jr, Graydon L [ORNL

    2013-11-01

    Fluoride salt-cooled High-temperature Reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics, and fully passive safety. This roadmap describes the principal remaining FHR technology challenges and the development path needed to address the challenges. This roadmap also provides an integrated overview of the current status of the broad set of technologies necessary to design, evaluate, license, construct, operate, and maintain FHRs. First-generation FHRs will not require any technology breakthroughs, but do require significant concept development, system integration, and technology maturation. FHRs are currently entering early phase engineering development. As such, this roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant, the lack of an approved licensing framework, the lack of qualified, salt-compatible structural materials, and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

  13. Concept of an inherently-safe high temperature gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro [Nuclear Hydrogen and Heat Application Research Center, Japan Atomic Energy Agency, Oarai-machi, Ibaraki-ken, 311-1394 (Japan)

    2012-06-06

    As the challenge to ensure no harmful release of radioactive materials at the accidents by deterministic approach instead to satisfy acceptance criteria or safety goal for risk by probabilistic approach, new concept of advanced reactor, an inherently-safe high temperature gas-cooled reactor, is proposed based on the experience of the operation of the actual High Temperature Gas-cooled Reactor (HTGR) in Japan, High Temperature Engineering Test Reactor (HTTR), and the design of the commercial plant (GTHTR300), utilizing the inherent safety features of the HTGR (i.e., safety features based on physical phenomena). The safety design philosophy of the inherently-safe HTGR for the safety analysis of the radiological consequences is determined as the confinement of radioactive materials is assured by only inherent safety features without engineered safety features, AC power or prompt actions by plant personnel if the design extension conditions occur. Inherent safety features to prevent the loss or degradation of the confinement function are identified. It is proposed not to apply the probabilistic approach for the evaluation of the radiological consequences of the accidents in the safety analysis because no inherent safety features fail for the mitigation of the consequences of the accidents. Consequently, there are no event sequences to harmful release of radioactive materials if the design extension conditions occur in the inherently-safe HTGR concept. The concept and future R and D items for the inherently-safe HTGR are described in this paper.

  14. Emergency reactor core cooling water injection device for light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oda, Junro.

    1994-05-13

    A reactor pressure vessel is immersed in pool water of a reactor container. A control valve is interposed to a water supplying pipelines connecting pool water and a pressure vessel. A valve actuation means for opening/closing the control valve comprises a lifting tank. The inner side of the lifting tank and the inner side of the pressure vessel are connected by a communication pipeline (a syphon pipe) at upper and lower two portions. The lifting tank and the control valve are connected by a link mechanism. When a water level in the pressure vessel is lowered, the water level in the lifting tank is lowered to the same level as that in the pressure vessel. This reduces the weight of the lifting tank, the lifting tank is raised, to open the control valve by way of a link mechanism. As a result, liquid phase in the pressure vessel is in communication with the pool water, and the pool water flows down into the pressure vessel to maintain the reactor core in a flooded state. (I.N.).

  15. Development of level-1 PSA method applicable to Japan Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kurisaka, K., E-mail: kurisaka.kennichi@jaea.go.jp [Advanced Nuclear System R and D Directorate, Japan Atomic Energy Agency, Ibaraki (Japan); Sakai, T.; Yamano, H. [Advanced Nuclear System R and D Directorate, Japan Atomic Energy Agency, Ibaraki (Japan); Fujita, S.; Minagawa, K. [Department of Mechanical Engineering, School of Engineering, Tokyo Denki University, Tokyo (Japan); Yamaguchi, A.; Takata, T. [Department of Energy and Environment Engineering, Osaka University, Osaka (Japan)

    2014-04-01

    This paper describes a study to develop the level-1 probabilistic safety assessment (PSA) method that is applicable to the Japan Sodium-cooled Fast Reactor (JSFR). This study has been started since August 2010 and aims to provide a new evaluation method of (1) passive safety architectures related to internal events and (2) an advanced seismic isolation system related to a seismic event as a representative external event in Japan. Regarding the internal events evaluation, a quantitative analysis on the frequency of the core damage caused by reactor shutdown failure was conducted. A failure in passive reactor shutdown was taken into account in the event tree model. The failure rate of sodium-cooled fast reactor (SFR) specific components was evaluated based on the operating experience in existing SFRs by applying the Hierarchical Bayesian Method, which can consider a plant-to-plant variability. By conducting an uncertainty analysis, it was found that the assumption about the correlation of the probability parameters between the main and backup reactor shutdown systems (RSSs) is sensitive to the mean value of the frequency of the core damage caused by reactor shutdown failure. As for the seismic event evaluation, seismic response analysis and sensitivity analysis of a seismic isolation system were carried out. Rubber bearings have a hardening property in horizontal direction and a softening property in vertical direction in case of large deformation. Therefore the analyses considered nonlinearity of rubber bearings. Both horizontal and vertical nonlinear characteristics of rubber bearings were explained by multi-linear model. Mass point analytical models were applied. At first, seismic response analysis was executed in order to investigate influence of nonlinearity of rubber bearing upon response of building. Then sensitivity analysis was executed. Parameters of rubber bearings, oil dampers and the building were fluctuated, and influence of dispersion of these

  16. Experimental investigation of a directionally enhanced DHX concept for high temperature Direct Reactor Auxiliary Cooling Systems

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, Joel T.; Blandford, Edward D., E-mail: edb@unm.edu

    2016-07-15

    Highlights: • A novel directional heat exchanger design has been developed. • Hydrodynamic tests have been performed on the proposed design. • Heat transfer performance is inferred by hydrodynamic results. • Results are discussed and future work is suggested. - Abstract: The use of Direct Reactor Auxiliary Cooling Systems (DRACSs) as a safety-related decay heat removal system for advanced reactors has developed historically through the Sodium Fast Reactor (SFR) community. Beginning with the EBR-II, DRACSs have been utilized in a large number of past and current SFR designs. More recently, the DRACS has been adopted for Fluoride Salt-Cooled High-Temperature Reactors (FHRs) for similar decay heat removal functions. In this paper we introduce a novel directionally enhanced DRACS Heat Exchanger (DHX) concept. We present design options for optimizing such a heat exchanger so that shell-side heat transfer is enhanced in one primary coolant flow direction and degraded in the opposite coolant flow direction. A reduced-scale experiment investigating the hydrodynamics of a directionally enhanced DHX was built and the data collected is presented. The concept of thermal diodicity is expanded to heat exchanger technologies and used as performance criteria for evaluating design options. A heat exchanger that can perform as such would be advantageous for use in advanced reactor concepts where primary coolant flow reversal is expected during Loss-of-Forced-Circulation (LOFC) accidents where the ability to circulate coolant is compromised. The design could also find potential use in certain advanced Sodium Fast Reactor (SFR) designs utilizing fluidic diode concepts.

  17. Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hoon; Chang, Soon Heung; Choi, Tae Hoon; Kim, Hyun Sop; Yang, Soo Hyung; Kim, Soo Hyung; Kim, Seung Hop; An Hyung Taek; Jeong, Yong Hoon; Huh, Gyun Young [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-03-15

    Cooling methodologies for the molten corium resulted from the severe accident of the nuclear power plant is suggested as one of most important items for the safety of the NPP. In this regard, considerable experimental and analytical works have been devoted. In the 1st phase of this project, present status related to the external reactor vessel cooling for the retention of the corium in the reactor vessel and corium at the reactor cavity have been investigated and preliminary studies have been accomplished for the detail evaluation of the each cooling methodology. The preliminary studies include the analysis and detail investigation of the possible phenomena, investigation of the heat transfer correlations and preliminary evaluation of the external reactor vessel cooling using the developed computer code.

  18. Reliability and safety of the K Reactor cooling system: Part 2, Engineering analysis of hydraulic and mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Shoemaker, R.H.

    1960-04-04

    Subsequent to the recent formulation and adoption of safety criteria for reactor cooling systems, there appeared the need for an independent evaluation of the safety and reliability of the K-Reactor cooling system in terms of these criteria. The primary, secondary and last-ditch cooling systems of this reactor involve a strong inter-dependence between electrical and hydraulic components of the water plant. Because of the complexity of inter-relationships between these components, the analysis was divided into two parallel studies which were accomplished during the simmer of 1959. F. D. Robbins has presented his analysis of the electrical power and control system in HW-61887. This report deals with an engineering analysis of the hydraulic and mechanical aspects of the reliability and safety of the K-Reactor Cooling System. The system, as described in this report, is that which existed during the simmer of 1959, prior to modification under Project CG-775 (now Project CG-883).

  19. A review of existing gas-cooled reactor circulators with application of the lessons learned to the new production reactor circulators

    Energy Technology Data Exchange (ETDEWEB)

    White, L.S.

    1990-07-01

    This report presents the results of a study of the lessons learned during the design, testing, and operation of gas-cooled reactor coolant circulators. The intent of this study is to identify failure modes and problem areas of the existing circulators so this information can be incorporated into the design of the circulators for the New Production Reactor (NPR)-Modular High-Temperature Gas Cooled Reactor (MHTGR). The information for this study was obtained primarily from open literature and includes data on high-pressure, high-temperature helium test loop circulators as well as the existing gas cooled reactors worldwide. This investigation indicates that trouble free circulator performance can only be expected when the design program includes a comprehensive prototypical test program, with the results of this test program factored into the final circulator design. 43 refs., 7 tabs.

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

    Science.gov (United States)

    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.

  1. Tritium permeation behavior through pyrolytic carbon in tritium production using high-temperature gas-cooled reactor for fusion reactors

    Directory of Open Access Journals (Sweden)

    H. Ushida

    2016-12-01

    Full Text Available Under tritium production method using a high-temperature gas-cooled reactor loaded Li compound, Li compound has to be coated by ceramic materials in order to suppress the spreading of tritium to the whole reactor. Pyrolytic carbon (PyC is a candidate of the coating material because of its high resistance for gas permeation. In this study, hydrogen permeation experiments using a PyC-coated isotropic graphite tube were conducted and hydrogen diffusivity, solubility and permeability were evaluated. Tritium permeation behavior through PyC-coated Li compound particles was simulated by using obtained data. Hydrogen permeation flux through PyC in a steady state is proportional to the hydrogen pressure and is larger than that through Al2O3 which is also candidate coating material. However, total tritium leak within the supposed reactor operation period through the PyC-coated Li compound particles is lower than that through the Al2O3-coated ones because the hydrogen absorption capacity in PyC is considerably larger than that in Al2O3.

  2. Nonlinear Dynamic Modeling and Simulation of a Passively Cooled Small Modular Reactor

    Science.gov (United States)

    Arda, Samet Egemen

    A nonlinear dynamic model for a passively cooled small modular reactor (SMR) is developed. The nuclear steam supply system (NSSS) model includes representations for reactor core, steam generator, pressurizer, hot leg riser and downcomer. The reactor core is modeled with the combination of: (1) neutronics, using point kinetics equations for reactor power and a single combined neutron group, and (2) thermal-hydraulics, describing the heat transfer from fuel to coolant by an overall heat transfer resistance and single-phase natural circulation. For the helical-coil once-through steam generator, a single tube depiction with time-varying boundaries and three regions, i.e., subcooled, boiling, and superheated, is adopted. The pressurizer model is developed based upon the conservation of fluid mass, volume, and energy. Hot leg riser and downcomer are treated as first-order lags. The NSSS model is incorporated with a turbine model which permits observing the power with given steam flow, pressure, and enthalpy as input. The overall nonlinear system is implemented in the Simulink dynamic environment. Simulations for typical perturbations, e.g., control rod withdrawal and increase in steam demand, are run. A detailed analysis of the results show that the steady-state values for full power are in good agreement with design data and the model is capable of predicting the dynamics of the SMR. Finally, steady-state control programs for reactor power and pressurizer pressure are also implemented and their effect on the important system variables are discussed.

  3. Heat transfer and core neutronics considerations of the heat pipe cooled thermionic reactor

    Science.gov (United States)

    Determan, W. R.; Lewis, Brian

    1991-01-01

    The authors summarize the results of detailed neutronic and thermal-hydraulic evaluations of the heat pipe cooled thermionic (HPTI) reactor design, identify its key design attributes, and quantify its performance characteristics. The HPTI core uses modular, liquid-metal core heat transfer assemblies to replace the liquid-metal heat transport loop employed by in-core thermionic reactor designs of the past. The nuclear fuel, power conversion, heat transport, and heat rejection functions are all combined into a single modular unit. The reactor/converter assembly uses UN fuel pins to obtain a critical core configuration with in-core safety rods and reflector controls added to complete the subassembly. By thermally bonding the core heat transfer assemblies during the reactor core is coupled neutronically, thermally, and electrically into a modular assembly of individual power sources with cross-tied architecture. A forward-facing heat pipe radiator assembly extends from the reactor head in the shape of a frustum of a cone on the opposite side of the power system from the payload. Important virtues of the concept are the absence of any single-point failures and the ability of the core to effectively transfer the TFE waste heat load laterally to other in-core heat transfer assemblies in the event of multiple failures in either in-core and radiator heat pipes.

  4. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

  5. Heat transfer and core neutronics considerations of the heat pipe cooled thermionic reactor

    Science.gov (United States)

    Determan, W. R.; Lewis, Brian

    The authors summarize the results of detailed neutronic and thermal-hydraulic evaluations of the heat pipe cooled thermionic (HPTI) reactor design, identify its key design attributes, and quantify its performance characteristics. The HPTI core uses modular, liquid-metal core heat transfer assemblies to replace the liquid-metal heat transport loop employed by in-core thermionic reactor designs of the past. The nuclear fuel, power conversion, heat transport, and heat rejection functions are all combined into a single modular unit. The reactor/converter assembly uses UN fuel pins to obtain a critical core configuration with in-core safety rods and reflector controls added to complete the subassembly. By thermally bonding the core heat transfer assemblies during the reactor core is coupled neutronically, thermally, and electrically into a modular assembly of individual power sources with cross-tied architecture. A forward-facing heat pipe radiator assembly extends from the reactor head in the shape of a frustum of a cone on the opposite side of the power system from the payload. Important virtues of the concept are the absence of any single-point failures and the ability of the core to effectively transfer the TFE waste heat load laterally to other in-core heat transfer assemblies in the event of multiple failures in either in-core and radiator heat pipes.

  6. Conceptual System Design of a Supercritical CO2 cooled Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seong Gu; Oh, Bongseong; Baik, Seung Joon; Yu, Hwanyeal; Kim, Yonghee; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The S-CO2 Brayton cycle has many advantages for SMR's power conversion system. The S-CO2 cycle can achieve small component size and simple cycle layout as shown in Fig. 1. Therefore, a concept of one module containing the S-CO2 cooled fast reactor core and power conversion system is realizable. Thanks to the compact heat exchanger technology such as Printed Circuit Heat Exchanger (PCHE), the supercritical fluid with mediocre heat transfer performance can be utilized to a thermal cycle. This concept of fully modularized reactor is named as KAIST Micro Modular Reactor (MMR). It can achieve large economic by production in series, and transported in the land way or sea way. Based on the design results and dimensions of the reactor core and cycle components, the authors propose a conceptual layout of KAIST MMR. Based on this concept of reactor core, power conversion system, and decay heat removal system, the seasonal operation and transient analysis will be performed in the further works.

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sujin; Kim, Jeongeun; Kim, Hyeonil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

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

  8. Validation of SCALE for High Temperature Gas-Cooled Reactors Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Ilas, Dan [ORNL; Kelly, Ryan P [ORNL; Sunny, Eva E [ORNL

    2012-08-01

    This report documents verification and validation studies carried out to assess the performance of the SCALE code system methods and nuclear data for modeling and analysis of High Temperature Gas-Cooled Reactor (HTGR) configurations. Validation data were available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhE Handbook), prepared by the International Reactor Physics Experiment Evaluation Project, for two different HTGR designs: prismatic and pebble bed. SCALE models have been developed for HTTR, a prismatic fuel design reactor operated in Japan and HTR-10, a pebble bed reactor operated in China. The models were based on benchmark specifications included in the 2009, 2010, and 2011 releases of the IRPhE Handbook. SCALE models for the HTR-PROTEUS pebble bed configuration at the PROTEUS critical facility in Switzerland have also been developed, based on benchmark specifications included in a 2009 IRPhE draft benchmark. The development of the SCALE models has involved a series of investigations to identify particular issues associated with modeling the physics of HTGRs and to understand and quantify the effect of particular modeling assumptions on calculation-to-experiment comparisons.

  9. A Compact Gas-Cooled Fast Reactor with an Ultra-Long Fuel Cycle

    Directory of Open Access Journals (Sweden)

    Hangbok Choi

    2013-01-01

    Full Text Available In an attempt to allow nuclear power to reach its full economic potential, General Atomics is developing the Energy Multiplier Module (EM2, which is a compact gas-cooled fast reactor (GFR. The EM2 augments its fissile fuel load with fertile materials to enhance an ultra-long fuel cycle based on a “convert-and-burn” core design which converts fertile material to fissile fuel and burns it in situ over a 30-year core life without fuel supplementation or shuffling. A series of reactor physics trade studies were conducted and a baseline core was developed under the specific physics design requirements of the long-life small reactor. The EM2 core performance was assessed for operation time, fuel burnup, excess reactivity, peak power density, uranium utilization, etc., and it was confirmed that an ultra-long fuel cycle core is feasible if the conversion is enough to produce fissile material and maintain criticality, the amount of matrix material is minimized not to soften the neutron spectrum, and the reactor core size is optimized to minimize the neutron loss. This study has shown the feasibility, from the reactor physics standpoint, of a compact GFR that can meet the objectives of ultra-long fuel cycle, factory-fabrication, and excellent fuel utilization.

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

    Energy Technology Data Exchange (ETDEWEB)

    Oktamuliani, Sri, E-mail: srioktamuliani@ymail.com; Su’ud, Zaki, E-mail: szaki@fi.itb.ac.id [Nuclear and Reactor Physics Laboratory, FMIPA, ITB, Physics Buildings, Jl. Ganesha 10, Bandung 40132 (Indonesia)

    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.

  11. An Experimental Test Facility to Support Development of the Fluoride Salt Cooled High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoder Jr, Graydon L [ORNL; Aaron, Adam M [ORNL; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK); Fugate, David L [ORNL; Holcomb, David Eugene [ORNL; Kisner, Roger A [ORNL; Peretz, Fred J [ORNL; Robb, Kevin R [ORNL; Wilgen, John B [ORNL; Wilson, Dane F [ORNL

    2014-01-01

    The need for high-temperature (greater than 600 C) energy exchange and delivery systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The Fluoride Salt-Cooled High-Temperature Reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the Fluoride Salt Cooled High Temperature Reactor concept. The facility is capable of operating at up to 700 C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system; trace heating system; and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride salt heat transfer inside a heated pebble bed similar to that used for the core of the pebble bed advanced high-temperature reactor. This document describes the details of the loop design, auxiliary systems used to support the facility, the inductive heating system, and facility capabilities.

  12. THATCH: A computer code for modelling thermal networks of high- temperature gas-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.; Kennett, R.J.; Colman, J.; Ginsberg, T. (Brookhaven National Lab., Upton, NY (United States))

    1991-10-01

    This report documents the THATCH code, which can be used to model general thermal and flow networks of solids and coolant channels in two-dimensional r-z geometries. The main application of THATCH is to model reactor thermo-hydraulic transients in High-Temperature Gas-Cooled Reactors (HTGRs). The available modules simulate pressurized or depressurized core heatup transients, heat transfer to general exterior sinks or to specific passive Reactor Cavity Cooling Systems, which can be air or water-cooled. Graphite oxidation during air or water ingress can be modelled, including the effects of added combustion products to the gas flow and the additional chemical energy release. A point kinetics model is available for analyzing reactivity excursions; for instance due to water ingress, and also for hypothetical no-scram scenarios. For most HTGR transients, which generally range over hours, a user-selected nodalization of the core in r-z geometry is used. However, a separate model of heat transfer in the symmetry element of each fuel element is also available for very rapid transients. This model can be applied coupled to the traditional coarser r-z nodalization. This report described the mathematical models used in the code and the method of solution. It describes the code and its various sub-elements. Details of the input data and file usage, with file formats, is given for the code, as well as for several preprocessing and postprocessing options. The THATCH model of the currently applicable 350 MW{sub th} reactor is described. Input data for four sample cases are given with output available in fiche form. Installation requirements and code limitations, as well as the most common error indications are listed. 31 refs., 23 figs., 32 tabs.

  13. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1983-06-01

    During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

  14. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1983-06-01

    During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

  15. A dynamic model of a passively cooled small modular reactor for controller design purposes

    Energy Technology Data Exchange (ETDEWEB)

    Arda, Samet E., E-mail: s.e.arda@asu.edu; Holbert, Keith E., E-mail: holbert@asu.edu

    2015-08-15

    Highlights: • A mathematical dynamic model is developed for a passively cooled small modular reactor. • Reactor response associated single-phase natural circulation is analyzed. • A moving boundary model for a helical-coil steam generator is analyzed. • Dynamic responses of the overall model to representative perturbations are evaluated. • This compact model can be utilized for control system design. - Abstract: An analytical dynamic model for a passively cooled small modular reactor (SMR) is developed using a state-variable lumped parameter approach. Reactor power is represented by the generation time formulation of the point kinetics equations with a single combined neutron precursor group. The heat transfer process in the core is described via an overall heat transfer coefficient by defining two coolant lumps paired to a single fuel lump. In addition, a thermal–hydraulics model for single-phase natural circulation is incorporated. For the helical-coil steam generator, a moving-boundary model including subcooled, two-phase, and superheated regions is utilized. Finally, the hot leg riser and downcomer regions are expressed by first-order lags. The performance of the overall system described by ordinary differential equations (ODEs) is evaluated by the Simulink dynamic environment and directly using a MATLAB ODE solver recommended for stiff systems. Simulation results based on NuScale SMR design data show that the initial steady-state values for 100% power are within range of the design data and the model can predict the system dynamics due to typical perturbations, e.g., control rod movement and change in feedwater mass flow rate and temperature. The model developed in this work can be utilized as a foundation for designing and testing a suitable control algorithm for reactor thermal power.

  16. Development of GAMMA Code and Evaluation for a Very High Temperature gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Chang H; Lim, H.S.; Kim, E.S.; NO, H.C.

    2007-06-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. This paper will also include what improvements will be made in the Gamma code for the VHTR.

  17. Application of Gamma code coupled with turbomachinery models for high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2008-02-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-ofcoolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of a toxic gas, CO, and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. GAMMA code is being developed to implement turbomachinery models in the power conversion unit (PCU) and ultimately models associated with the hydrogen plant. Some preliminary results will be described in this paper.

  18. Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-02-01

    The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR

  19. The Role of Expansion and Fragmentation Phenomena on the Generation and Chemical Composition of Dust Particles in a Flash Converting Reactor

    Science.gov (United States)

    Duarte-Ruiz, Cirilo Andrés; Pérez-Tello, Manuel; Parra-Sánchez, Víctor Roberto; Sohn, Hong Yong

    2016-10-01

    A compositional fragmentation model was used to clarify the effect of expansion and fragmentation phenomena on the generation and chemical composition of dust particles in a flash converting reactor. A fragmentation index is introduced to represent the fraction of particles undergoing fragmentation, as opposed to expansion, within the particle population. Under typical operating conditions, the local dust content and the net amount of dust generated compared with the dust content in the feed first decreased and then increased along the reactor length, whereas the amount of particles undergoing fragmentation (fragmentation index) increased steadily. Dust generation was found to be the result of two competing phenomena, i.e., the expansion of dust particles in the feed and the production of dust from fragmentation of large particles. At short distance from the burner tip, the dust mostly consists of particles in the feed undergoing oxidation and expansion, whereas farther down the reactor it mostly consists of fragments of partially reacted particles. Based on the computer simulations under a variety of experimental conditions, a map of dust generation against fragmentation index was developed. For most practical purposes, dust generation may be approximated by the change in the mass fraction of dust in the population. At the reactor exit, the composition of the dust is approximately the same as the entire particle population.

  20. Gas-Cooled Fast Reactor: A Historical Overview and Future Outlook

    Directory of Open Access Journals (Sweden)

    W. F. G. van Rooijen

    2009-01-01

    Full Text Available A review is given of developments in the area of Gas-Cooled Fast Reactors (GCFR in the period from roughly 1960 until 1980. During that period, the GCFR concept was expected to increase the breeding gain, the thermal efficiency of a nuclear power plant, and alleviate some of the problems associated with liquid metal coolants. During this period, the GCFR concept was found to be more challenging than liquid-metal-cooled reactors, and none were ever constructed. In the second part of the paper, we provide an overview of the investigations on GCFR since the year 2000, when the Generation IV Initiative rekindled interest in this reactor type. The new GCFR concepts focus primarily on sustainable nuclear power, with very efficient resource use, minimum waste, and a very strong focus on (passive safety. An overview is presented of the main design characteristics of these Gen IV GCFRs, and a literature list is provided to guide the interested reader towards more detailed publications.

  1. Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

    Science.gov (United States)

    Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

    2017-01-01

    The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Directory of Open Access Journals (Sweden)

    Shixiong Song

    2014-01-01

    CFD results and empirical correlations’ predictions of pressure drop and local Nusselt numbers. Local pebble surface temperature distributions in several default conditions are investigated. Thermal removal capacities of molten salt are confirmed in the case of nominal condition; the pebble surface temperature under the condition of local power distortion shows the tolerance of pebble in extreme neutron dose exposure. The numerical experiments of local pebble insufficient cooling indicate that in the molten salt cooled pebble bed reactor, the pebble surface temperature is not very sensitive to loss of partial coolant. The methods and results of this paper would be useful for optimum designs and safety analysis of molten salt cooled pebble bed reactors.

  4. Comparison of sodium and lead-cooled fast reactors regarding reactor physics aspects, severe safety and economical issues

    Energy Technology Data Exchange (ETDEWEB)

    Tucek, Kamil [Joint Research Centre of the European Commission, Institute for Energy, Postbus 2, NL-1755 ZG Petten (Netherlands)]. E-mail: kamil.tucek@jrc.nl; Carlsson, Johan [Joint Research Centre of the European Commission, Institute for Energy, Postbus 2, NL-1755 ZG Petten (Netherlands); Wider, Hartmut [Joint Research Centre of the European Commission, Institute for Energy, Postbus 2, NL-1755 ZG Petten (Netherlands)

    2006-08-15

    A large number of new fast reactors may be needed earlier than foreseen in the Generation IV plans. According to the median forecast of the Special Report on Emission Scenarios commissioned by the Intergovernmental Panel on Climate Control nuclear power will increase by a factor of four by 2050. The drivers for this expected boost are the increasing energy demand in developing countries, energy security, but also climate concerns. However, staying with a once-through cycle will lead to both a substantially increased amount of high-level nuclear waste and an upward pressure on the price of uranium and even concerns about its availability in the coming decades. Therefore, it appears wise to accelerate the development of fast reactors and efficient re-processing technologies. In this paper, two fast reactor systems are discussed-the sodium-cooled fast reactor, which has already been built and can be further improved, and the lead-cooled fast reactor that could be developed relatively soon. An accelerated development of the latter is possible due to the sizeable experience on lead/bismuth eutectic coolant in Russian Alpha-class submarine reactors and the research efforts on accelerator-driven systems in the EU and other countries. First, comparative calculations on critical masses, fissile enrichments and burn-up swings of mid-sized SFRs and LFRs (600 MW{sub e}) are presented. Monte Carlo transport and burn-up codes were used in the analyses. Moreover, Doppler and coolant temperature and axial fuel expansion reactivity coefficients were also evaluated with MCNP and subsequently used in the European Accident Code-2 to calculate reactivity transients and unprotected Loss-of-Flow (ULOF) and Loss-of-Heat Sink (ULOHS) accidents. Further, ULOFs as well as decay heat removal (protected Total Loss-of-Power, TLOP) were calculated with the STAR-CD CFD code for both systems. We show that LFRs and SFRs can be used both as burners and as self-breeders, homogeneously incinerating

  5. Flow boiling CHF enhancement in an external reactor vessel cooling (ERVC) channel using graphene oxide nanofluid

    Energy Technology Data Exchange (ETDEWEB)

    Park, Seong Dae; Bang, In Cheol, E-mail: icbang@unist.ac.kr

    2013-12-15

    Highlights: • We investigate CHF limits of graphene oxide nanofluid for IVR-ERVC. • Graphene oxide nanofluid enhanced CHF up to about 20%. • CHF enhancement can be explained by the improved thermal activity. - Abstract: External reactor vessel cooling for in-vessel retention of corium is an important concept to mitigate the consequences of a severe accident by flooding the reactor cavity. Although this system has some merits, it is restricted by the capacity of heat removal through the nucleate boiling on the outer surface of the reactor. In this study, the graphene oxide (GO) nanofluid at 0.0001 vol% was used to enhance the critical heat flux (CHF). The CHF tests were conducted with a closed-loop facility. Test section simulated the reactor vessel of APR-1400 with a small scale. The test results show about ∼20% enhancement of CHF at 50 and 100 kg/m{sup 2} s under a 10 K subcooling condition. It means that the additional thermal margin could be acquired by just adding the GO nanoparticles to the flooding water without severe economic concerns. It is also found that this CHF enhancement is caused by coating the graphene oxide nanoparticles on the heated surface. However, the sessile drop tests on the coated heater surface show that the wettability of GO coated surface is not improved. The results of IR thermography show that one of the promising reasons is the change of thermal activity due to the coated GO nanoparticles on the heated surface.

  6. Design Study of Small Pb-Bi Cooled Modified Candle Reactors

    Science.gov (United States)

    Su'ud, Zaki; Sekimoto, H.

    2010-06-01

    In this study application of modified CANDLE burnup scheme based long life Pb-Bi Cooled Fast Reactors for small long life reactors with natural Uranium as Fuel Cycle Input has been performed. The reactor cores are subdivided into several parts with the same volume in the axial directions. The natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2, and 10 years after that it is shifted to region 3. This concept is applied to all regions, i.e. shifted the core of I'th region into I+1 region after the end of 10 years burn-up cycle. The first region 1 is filled by fresh natural uranium fuel. Compared to the previous works, in a smaller reactor core the criticality need to be considered more carefully especially at the beginning of life. As an optimized design, a core of 85 cm radius and 150 cm height with 300 MWt power are selected. This core can be operated 10 years without refueling or fuel shuffling. The average discharge burn-up is 350 GWd/ton HM.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  9. Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

    Directory of Open Access Journals (Sweden)

    Yeong-il Kim

    2013-01-01

    Full Text Available Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. CFD-DEM simulation of a conceptual gas-cooled fluidized bed nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, Lucilla C.; Su, Jian, E-mail: lucillalmeida@gmail.com, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Aguirre, Joao, E-mail: aguirre@rocky-dem.com [Engineering Simulation and Scientific Software (ESSS), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    Several conceptual designs of the fluidized-bed nuclear reactor have been proposed due to its many advantages over conventional nuclear reactors such as PWRs and BWRs. Amongst their characteristics, the enhanced heat transfer and mixing enables a more uniform temperature distribution, reducing the risk of hot-spot and excessive fuel temperature, in addition to resulting in a higher burnup of the fuel. Furthermore, the relationship between the bed height and reactor neutronics turns the coolant flow rate control into a power production mechanism. Moreover, the possibility of removing the fuel by gravity from the movable core in case of a loss-of-cooling accident increases its safety. High-accuracy modeling of particles and coolant flow in fluidized bed reactors is needed to evaluate reliably the thermal-hydraulic efficiency and safety margin. The two-way coupling between solid and fluid can account for high-fidelity solid-solid interaction and reasonable accuracy in fluid calculation and fluid-solid interaction. In the CFD-DEM model, the particles are modeled as a discrete phase, following the DEM approach, whereas the fluid flow is treated as a continuous phase, described by the averaged Navier-Stokes equations on a computational cell scale. In this work, the coupling methodology between Fluent and Rocky is described. The numerical approach was applied to the simulation of a bubbling fluidized bed and the results were compared to experimental data and showed good agreement. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    In, W. K.; Chun, T. H.; Lee, W. J.; Chang, J. H. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

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

  13. HYBRID SULFUR CYCLE FLOWSHEETS FOR HYDROGEN PRODUCTION USING HIGH-TEMPERATURE GAS-COOLED REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Gorensek, M.

    2011-07-06

    Two hybrid sulfur (HyS) cycle process flowsheets intended for use with high-temperature gas-cooled reactors (HTGRs) are presented. The flowsheets were developed for the Next Generation Nuclear Plant (NGNP) program, and couple a proton exchange membrane (PEM) electrolyzer for the SO2-depolarized electrolysis step with a silicon carbide bayonet reactor for the high-temperature decomposition step. One presumes an HTGR reactor outlet temperature (ROT) of 950 C, the other 750 C. Performance was improved (over earlier flowsheets) by assuming that use of a more acid-tolerant PEM, like acid-doped poly[2,2'-(m-phenylene)-5,5'-bibenzimidazole] (PBI), instead of Nafion{reg_sign}, would allow higher anolyte acid concentrations. Lower ROT was accommodated by adding a direct contact exchange/quench column upstream from the bayonet reactor and dropping the decomposition pressure. Aspen Plus was used to develop material and energy balances. A net thermal efficiency of 44.0% to 47.6%, higher heating value basis is projected for the 950 C case, dropping to 39.9% for the 750 C case.

  14. Comparison of containment systems for large sodium-cooled breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Seeman, S.E.; Armstrong, G.R.

    1978-04-01

    Three types of containment configurations for large breeder reactors (1000 MWe/2500 MWt) were compared to determine characteristic responses of each type to selected hypothetical accident conditions. The three types of system configurations were the single, double and containment-confinement designs. Two types of postulated hypothetical accidents were applied as a basis for evaluation--the energetic core disassambly accident and the whole core melt-through accident. More than 20 separate cases were analyzed in which design and accident assumptions were varied. Based on this study the containment-confinement configuration is shown to be the most promising for limiting off-site radiation doses for large sodium-cooled breeder reactors given the specific assumptions on the initial accident conditions.

  15. 2400MWt GAS-COOLED FAST REACTOR DHR STUDIES STATUS UPDATE.

    Energy Technology Data Exchange (ETDEWEB)

    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.

  16. Overall system description and safety characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Jae Woon; Chang, Jin Wook; Lim, Jae Yong; Cheon, Jin Sik; Lee, Tae Ho; Kim, Sung Kyun; Lee, Kwi Lim; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-08

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

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

    Science.gov (United States)

    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.

  19. Development of objective provision trees for Sodium-Cooled Fast Reactor Defense-in-depth evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Huichang [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    KALIMER is one of sodium-cooled fast reactor and being developed by Korea Atomic Energy Research Institute (KAERI), was developed and suggested in this paper. Developed OPT is for the defense-in-depth level 3, core heat removal safety function. Using OPT method, the evaluation of defense-in-depth implementation for the design features of KALIMER reactors were tried in this study. To utilize the design information of KALIMER, challenges in OPTs which are under development in this study, were identified based on the system physical boundaries. This approach make the identification of possible and postulated challenges much clear and this will be a benefit to further identification of provisions in KALIMER design. OPTs for other levels of defense-in-depth and other safety functions are under development.

  20. A water cooled, lithium lead breeding blanket for a DEMO fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Rieger, M.; Biggio, M.; Farfaletti-Casali, F.; Tominetti, S.; Wu, J.; Zucchetti, M. (Commission of the European Communities, Ispra (Italy). Joint Research Centre); Labbe, P.; Baraer, L.; Gervaise, G.; Giancarli, L.; Roze, M.; Severi, Y.; Quintric-Bossy, J. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France))

    1991-04-01

    The main features of a tritium breeding blanket for a Demonstration Power Reactor involving the eutectic Pb-17Li as liquid breeder and water as coolant are presented. The configuration of the blanket segments and breeder modules as well as their arrangement inside the reactor vacuum vessel are outlined. The main design aspects and the corresponding design limits are reviewed, namely those related to thermomechanics, neutronics, magneto-hydrodynamics, tritium permeation and recovery. First results of safety analysis, in particular those connected with the rupture of a coolant tube in the breeder module are presented and discussed. As a conclusion, the feasibility of the concept look attractive. A problem which requires further investigation is that of the tritium self-sufficiency. It is shown that a net tritium production near to one can be obtained if berylium tiles are placed in front of the plasma, provided that they are cooled by heavy water. (orig.).

  1. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) for Power and Process Heat

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, Charles [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Peterson, Per [Univ. of California, Berkeley, CA (United States); Sridharan, Kumar [Univ. of Wisconsin, Madison, WI (United States)

    2015-01-21

    In 2011 the U.S. Department of Energy through its Nuclear Energy University Program (NEUP) awarded a 3- year integrated research project (IRP) to the Massachusetts Institute of Technology (MIT) and its partners at the University of California at Berkeley (UCB) and the University of Wisconsin at Madison (UW). The IRP included Westinghouse Electric Company and an advisory panel chaired by Regis Matzie that provided advice as the project progressed. The first sentence of the proposal stated the goals: The objective of this Integrated Research Project (IRP) is to develop a path forward to a commercially viable salt-cooled solid-fuel high-temperature reactor with superior economic, safety, waste, nonproliferation, and physical security characteristics compared to light-water reactors. This report summarizes major results of this research.

  2. Minor actinides impact on basic safety parameters of medium-sized sodium-cooled fast reactor

    Directory of Open Access Journals (Sweden)

    Darnowski Piotr

    2015-03-01

    Full Text Available An analysis of the influence of addition of minor actinides (MA to the fast reactor fuel on the most important safety characteristics was performed. A special emphasis was given to the total control rods worth in order to describe qualitatively and quantitatively its change with MA content. All computations were performed with a homogeneous assembly model of modified BN-600 sodium-cooled fast reactor core with 0, 3 and 6% of MA. A model was prepared for the Monte Carlo neutron transport code MCNP5 for fresh fuel in the beginning-of-life (BOL state. Additionally, some other parameters, such as Doppler constant, sodium void reactivity, delayed neutron fraction, neutron fluxes and neutron spectra distribution, were computed and their change with MA content was investigated. Study indicates that the total control rods worth (CRW decreases with increasing MA inventory in the fuel and confirms that the addition of MA has a negative effect on the delayed neutron fraction.

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

    Energy Technology Data Exchange (ETDEWEB)

    Monado, Fiber [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su' ud, Zaki; Waris, Abdul; Basar, Khairul [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

    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.

  4. A small high temperature gas cooled reactor for nuclear marine propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Brugiere, F.; Sillon, C. [Ecole des Applications Militaires de l' Energie Atomique, 50 - Cherbourg (France); Foster, A.; Hamilton, P.; Jewer, S.; Thompson, A.C. [Defence College of Electromechanical Engineering, Nuclear Dept., Military Rd, Gosport (United Kingdom); Kingston, T.; Williams, A.M.; Beeley, P.A. [Rolls-Royce (Marine Power), Raynesway, Derby (United Kingdom)

    2007-07-01

    Results from a design study for a hypothetical nuclear marine propulsion plant are presented. The plant utilizes a small High Temperature Gas Cooled Reactor (HTGCR) similar to the GTHTR300 design by the Japan Atomic Energy Agency with power being generated by a direct cycle gas turbine. The GTHTR300 design is modified in order to achieve the required power of 80 MWth and core lifetime of approximately 10 years. Thermal hydraulic analysis shows that in the event of a complete loss of flow accident the hot channel fuel temperature exceeds the 1600 Celsius degrees limit due to the high power peaking in assemblies adjacent to the inner reflector. Reactor dynamics shows oscillatory behaviour in rapid power transients. An automatic control rod system is suggested to overcome this problem. (authors)

  5. Corrosion mechanisms of candidate structural materials for supercritical water-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    Lefu ZHANG; Fawen ZHU; Rui TANG

    2009-01-01

    Nickel-based alloys, austenitic stainless steel, ferritic/martensitic heat-resistant steels, and oxide dispersion strengthened steel are presently considered to be the candidate structural or fuel-cladding materials for supercritical water-cooled reactor (SCWR), one of the promising generation IV reactor for large-scale electric power production. However, corrosion and stress corrosion cracking of these candidate alloys still remain to be a major problem in the selection of nuclear fuel cladding and other structural materials, such as water rod. Survey of literature and experimental results reveal that the general corrosion mechanism of those candidate materials exhibits quite complicated mechanism in high-temperature and high-pressure supercritical water. Formation of a stable protective oxide film is the key to the best corrosion-resistant alloys. This paper focuses on the mechanism of corrosion oxide film breakdown for SCWR candidate materials.

  6. A Conceptual Study of a Supercritical CO2-Cooled Micro Modular Reactor

    Directory of Open Access Journals (Sweden)

    Hwanyeal Yu

    2015-12-01

    Full Text Available A neutronics conceptual study of a supercritical CO2-cooled micro modular reactor (MMR has been performed in this work. The suggested MMR is an extremely compact and truck-transportable nuclear reactor. The thermal power of the MMR is 36.2 MWth and it is designed to have a 20-year lifetime without refueling. A salient feature of the MMR is that all the components including the generator are integrated in a small reactor vessel. For a minimal volume and long lifetime of the MMR core, a fast neutron spectrum is utilized in this work. To enhance neutron economy and maximize the fuel volume fraction in the core, a high-density uranium mono-nitride U15N fuel is used in the fast-spectrum MMR. Unlike the conventional supercritical CO2-cooled fast reactors, a replaceable fixed absorber (RFA is introduced in a unique way to minimize the excess reactivity and the power peaking factor of the core. For a compact core design, the drum-type control absorber is adopted as the primary reactivity control mechanism. In this study, the neutronics analyses and depletions have been performed by using the continuous energy Monte Carlo Serpent code with the evaluated nuclear data file ENDF/B-VII.1 Library. The MMR core is characterized in view of several important safety parameters such as control system worth, fuel temperature coefficient (FTC and coolant void reactivity (CVR, etc. In addition, a preliminary thermal-hydraulic analysis has also been performed for the hottest channel of the Korea Advanced Institute of Science and Technology (KAIST MMR.

  7. High Temperature Gas-cooled Reactor Projected Markets and Scoping Economics

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2010-08-01

    The NGNP Project has the objective of developing the high temperature gas-cooled reactor (HTGR) technology to supply high temperature process heat to industrial processes as a substitute for burning of fossil fuels, such as natural gas. Applications of the HTGR technology that have been evaluated by the NGNP Project for supply of process heat include supply of electricity, steam and high-temperature gas to a wide range of industrial processes, and production of hydrogen and oxygen for use in petrochemical, refining, coal to liquid fuels, chemical, and fertilizer plants.

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

    OpenAIRE

    Zhu Guifeng; Zou Yang; Xu Hongjie

    2016-01-01

    Sustainability of thorium fuel in a Pebble-Bed Fluoride salt-cooled High temperature Reactor (PB-FHR) is investigated to find the feasible region of high discharge burnup and negative Flibe (2LiF-BeF2) salt Temperature Reactivity Coefficient (TRC). Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tristructural-isotropic (TRISO) coated particle system for increasing fuel loading and decreasing excessive moderation. To analyze the neutronic characteristics,...

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

    OpenAIRE

    Chan Bock Lee; Jin Sik Cheon; Sung Ho Kim; Jeong-Yong Park; Hyung-Kook Joo

    2016-01-01

    Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU)–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochem...

  10. Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sieger, Matt [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Moe, Wayne [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); HolbrookINL, Mark [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    The goal of this review is to enable application of codes or software packages for safety assessment of advanced sodium-cooled fast reactor (SFR) designs. To address near-term programmatic needs, the authors have focused on two objectives. First, the authors have focused on identification of requirements for software QA that must be satisfied to enable the application of software to future safety analyses. Second, the authors have collected best practices applied by other code development teams to minimize cost and time of initial code qualification activities and to recommend a path to the stated goal.

  11. A STRONGLY COUPLED REACTOR CORE ISOLATION COOLING SYSTEM MODEL FOR EXTENDED STATION BLACK-OUT ANALYSES

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Haihua [Idaho National Laboratory; Zhang, Hongbin [Idaho National Laboratory; Zou, Ling [Idaho National Laboratory; Martineau, Richard Charles [Idaho National Laboratory

    2015-03-01

    The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup cooling water to the reactor pressure vessel (RPV) when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. The RCIC system operates independently of AC power, service air, or external cooling water systems. The only required external energy source is from the battery to maintain the logic circuits to control the opening and/or closure of valves in the RCIC systems in order to control the RPV water level by shutting down the RCIC pump to avoid overfilling the RPV and flooding the steam line to the RCIC turbine. It is generally considered in almost all the existing station black-out accidents (SBO) analyses that loss of the DC power would result in overfilling the steam line and allowing liquid water to flow into the RCIC turbine, where it is assumed that the turbine would then be disabled. This behavior, however, was not observed in the Fukushima Daiichi accidents, where the Unit 2 RCIC functioned without DC power for nearly three days. Therefore, more detailed mechanistic models for RCIC system components are needed to understand the extended SBO for BWRs. As part of the effort to develop the next generation reactor system safety analysis code RELAP-7, we have developed a strongly coupled RCIC system model, which consists of a turbine model, a pump model, a check valve model, a wet well model, and their coupling models. Unlike the traditional SBO simulations where mass flow rates are typically given in the input file through time dependent functions, the real mass flow rates through the turbine and the pump loops in our model are dynamically calculated according to conservation laws and turbine/pump operation curves. A simplified SBO demonstration RELAP-7 model with this RCIC model has been successfully developed. The demonstration model includes the major components for the primary system of a BWR, as well as the safety

  12. Performance of materials in the component cooling water systems of pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B.S.

    1993-06-01

    The component cooling water (CCW) system provides cooling water to several important loads throughout the plant under all operating conditions. An aging assessment CCW systems in pressurized water reactors (PWRs) was conducted as part of Nuclear Plant Aging Research Program (NPAR) instituted by the US Nuclear Regulatory Commission. This paper presents some of the results on the performances of materials in respect of their application in CCW Systems. All the CCW system failures reported to the Nuclear Plant Reliability Data System (NPRDS) from January 1988 to June 1990 were reviewed; it is concluded that three of the main contributors to CCW system failures are valves, pumps, and heat exchangers. This study identified the modes and causes of failure for these components; most of the causes for the aging-related failures could be related to the performance of materials. Also, in this paper the materials used for these components are reviewed, and there aging mechanisms under CCW system conditions are discussed.

  13. Synthesis of ZnO particles in a quench-cooled flame reactor

    DEFF Research Database (Denmark)

    Hansen, Jens Peter; Jensen, Joakim Reimer; Livbjerg, Hans

    2001-01-01

    . At the highest tested production rate, the specific surface area of the ZnO particles increases from 20 to 60 m(2)/g when quenching is employed. The particles are characterized by BET surface area measurements, TEM images, and the size distributions of particle aggregates are measured by a scanning mobility......The quench cooling of a flame by injection of cold air was studied in a flame reactor for the formation of ZnO particles in a premixed flame with a precursor jet. A rapid temperature drop downstream from the temperature peak is advantageous for the attainment of a large specific surface area....... Computational fluid dynamics simulations were used to design a quench ring with nozzles directed slightly upward and at a small tangential angle from the direct line to the center. This novel design avoids distortion of the flow pattern below the quenching plane and effectively cools the flame immediately above...

  14. SPARC fast reactor design : Design of two passively safe metal-fuelled sodium-cooled pool-type small modular fast reactors with Autonomous Reactivity Control

    OpenAIRE

    Lindström, Tobias

    2015-01-01

    In this master thesis a small modular sodium-cooled metal-fuelled pool-type fast reactor design, called SPARC - Safe and Passive with Autonomous Reactivity control, has been designed. The long term reactivity changes in the SPARC are managed by implementation of the the Autonomous Reactivity Control (ARC) system, which is the novelty of the design. The overall design is mainly based on the Integral Fast Reactor project (IFR), which experimentally demonstrated the passive safety characteristic...

  15. Recycling option search for a 600-MWe sodium-cooled transmutation fast reactor

    Directory of Open Access Journals (Sweden)

    Yong Kyo Lee

    2015-02-01

    Full Text Available Four recycling scenarios involving pyroprocessing of spent fuel (SF have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR, KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs. If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE isotopes. The RE isotope recovery factor should be lowered to ≤20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

  16. The Effect of Duct Level on the Performance of Reactor Vault Cooling System in the PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Yeom, Sujin; Ryu, Seung Ho; Kim, Dehee; Lee, Tae-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Development of the prototype gen-Ⅵ sodium-cooled fast reactor (PGSFR) has been ongoing in Korea Atomic Energy Research Institute (KAERI). A reactor vault cooling system (RVCS), one of passive decay heat removal systems (PDHRS), passively removes core decay heat by chimney effect when severe accidents occur. The air cooling path is located around containment vessel (CV). An air separator which divides the downstream air and the upstream air is installed between CV and the concrete wall. To design the RVCS, key design parameters such as stack height, gap size between the concrete wall and the air separator, gap size between the air separator and the CV, thickness and layer composition of the air separator have to be determined. A duct level is one of these design parameters. It denotes the height of the upstream air path and related to the heat transfer length from CV to air. The duct level should be optimized with considering structural reliability and heat removal performance. Thus, in this paper, the heat removal performance of RVCS is evaluated depends on the duct level using 1D system design code, that is developed by KAERI autonomously, and commercial CFD program for optimum design of RVCS In this paper, the heat removal performance of RVCS is evaluated depends on the duct level using PARS2- LMR code and commercial CFD program for optimum design of RVCS to satisfy both conflicting needs, structural reliability and cooling performance. As a result of PARS2-LMR code analysis, it was observed that the heat removal rate increases as increase of duct level and the geometrical conditions, that satisfy the design limitations, were obtained. To qualitatively observe the trends of local temperature distribution, CFD simulations were conducted and hotspots were observed at the upper region of ducts for the low duct level case.

  17. ANALYSIS OF A HIGH TEMPERATURE GAS-COOLED REACTOR POWERED HIGH TEMPERATURE ELECTROLYSIS HYDROGEN PLANT

    Energy Technology Data Exchange (ETDEWEB)

    M. G. McKellar; E. A. Harvego; A. M. Gandrik

    2010-11-01

    An updated reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322°C and 750°C, respectively. The reactor heat is used to produce heat and electric power to the HTE plant. A Rankine steam cycle with a power conversion efficiency of 44.4% was used to provide the electric power. The electrolysis unit used to produce hydrogen includes 1.1 million cells with a per-cell active area of 225 cm2. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 42.8% at a hydrogen production rate of 1.85 kg/s (66 million SCFD) and an oxygen production rate of 14.6 kg/s (33 million SCFD). An economic analysis of this plant was performed with realistic financial and cost estimating The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.03/kg of hydrogen was calculated assuming an internal rate of return of 10% and a debt to equity ratio of 80%/20% for a reactor cost of $2000/kWt and $2.41/kg of hydrogen for a reactor cost of $1400/kWt.

  18. Neutronic assessment of liquid-metal cooled fast reactors using thorium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pilarski, Stevan [Electricite de France R et D, 1 Avenue du General de Gaulle, 92141 Clamart (France); Institut de Physique Nucleaire d' Orsay, 15 rue Georges Clemenceau 91406 Orsay (France)

    2009-06-15

    The long-term sustainability of atomic fission energy will require the development of new types of reactors, able to exceed the limits of the existing ones in terms of optimal use of natural resources, which clearly necessitates breeding of fissile material. In this context, fast reactors using uranium-plutonium fuel are the most mature solution from an industrial viewpoint. In addition to the obvious interest in terms of fuel resources, there is a major incentive to consider the use of the {sup 232}Th- {sup 233}U fuel cycle as an alternative to the traditional {sup 238}U-{sup 239}Pu cycle for fast reactors: it is an effective way of addressing the safety issue of the highly positive void reactivity effect, which is a well-known problem for liquid-metal cooled fast reactors of commercial size [1]. This work investigates the performance of liquid-metal cooled fast reactors in {sup 232}Th-{sup 233}U fuel cycle and draws a comparison with the traditional {sup 238}U-{sup 239}Pu cycle. Four coolants have been considered: Na, Pb, Mg(17%at.)-Pb and Li(17%at.)-Pb; a simulation of their use in cores ranging from 700 MWth to 3600 MWth has been performed in two-dimensional diffusion theory using the European system of codes ERANOS [2,3] developed at CEA. The performance parameters such as the breeding ratio have been computed for each concept, alongside safety-related parameters: the delayed neutron fraction, the cycle reactivity swing, the Doppler constant and other thermal feedbacks. More specifically, the issue of void reactivity is studied in detail using perturbation theory. These calculations are performed at equilibrium fuel composition and are complemented by the study of the initial fuel loading at start-up which is a mixture of {sup 232}Th-{sup 239}Pu. The isotopic composition of the fissile corresponds to the plutonium available from French reactors in 2035. The conclusions of this work are that near-zero to large negative void reactivity effects can be achieved in

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chung-yong; Lee, Sung-yong; Choi, Min-young; Lee, Seung-jae; Jo, Young-ho [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of); Lee, Young-woo; Cho, Moon-sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

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

  20. An experimental study of hypervapotron structure in external reactor vessel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yufeng; Zhang, Ming [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Hou, Fangxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); Gao, Tianfang [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Chen, Peipei, E-mail: chenpeipei@snptc.com.cn [State Power Investment Group Corporation, Beijing (China)

    2016-07-15

    Highlights: • Experiments are performed to study the application of hypervapotron in ERVC design. • CHF experiments on two surfaces are conducted under different flow conditions. • Hypervapotron improves CHF performance by 40–60% compared with smooth surface. • Visualization shows fin structure removes vapor mushroom for better liquid supply. - Abstract: In vessel retention (IVR) is one of the key strategies for many advanced LWR designs to mitigate postulated severe accidents. The success of IVR substantially relies on external reactor vessel cooling (ERVC) by which the decay heat is removed from the melt core in the reactor vessel lower head. The main challenge of IVR is to provide an adequate safety margin of ERVC against critical heat flux (CHF) of subcooled flow boiling in the reactor lower head flow channel. Due to uncertainties in corium melt pool configuration, large CHF margin of ERVC is usually required by regulatory authorities to demonstrate reliability of severe accident mitigation methods. Various CHF enhancement designs have been proposed and studied in literature. In this paper, an experimental study of hypervapotron structure as a novel design to improve CHF performance of ERVC is conducted. Hypervapotron is chosen as one of the potential engineering options for International Thermonuclear Experimental Reactor (ITER) program as a divertor structure to remove highly intense heat from fusion chamber. This study is to conduct CHF experiments at typical PWR ERVC working conditions. The CHF experiments are performed in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. Both smooth and hypervapotron surface are tested at various inclination angles of the test section to simulate various positions of the reactor lower head. The hypervapotron is found to have a 40–60% CHF improvement compared with the smooth surface. The high speed visualization indicates that hypervapotron is able to

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

    Science.gov (United States)

    Ilham, Muhammad; Su’ud, Zaki

    2017-01-01

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

  2. Piping Flexibility Analysis of the Primary Cooling System of TRIGA 2000 Bandung Reactor due to Earthquake

    Directory of Open Access Journals (Sweden)

    H.P. Rahardjo

    2011-08-01

    Full Text Available Earthquakes in a nuclear installation can overload a piping system which is not flexible enough. These loads can be forces, moments and stresses working on the pipes or equipments. If the load is too large and exceed the allowable limits, the piping and equipment can be damaged and lead to overall system operation failure. The load received by piping systems can be reduced by making adequate piping flexibility, so all the loads can be transmitted homogenously throughout the pipe without load concentration at certain point. In this research the analysis of piping stress has been conducted to determine the size of loads that occured in the piping of primary cooling system of TRIGA 2000 Reactor, Bandung if an earthquake happened in the reactor site. The analysis was performed using Caesar II software-based finite element method. The ASME code B31.1 arranging the design of piping systems for power generating system (Power Piping Code was used as reference analysis method. Modeling of piping systems was based on the cooling piping that has already been installed and the existing data reported in Safety Analysis Reports (SARs of TRIGA 2000 reactor, Bandung. The quake considered in this analysis is the earthquake that occurred due to the Lembang fault, since it has the Peak Ground Acceleration (PGA in the Bandung TRIGA 2000 reactor site. The analysis results showed that in the static condition for sustain and expansion loads, the stress fraction in all piping lines does not exceed the allowable limit. However, during operation moment, in dynamic condition, the primary cooling system is less flexible at sustain load, ekspansi load, and combination load and the stress fraction have reached 95,5%. Therefore a pipeline modification (rerouting is needed to make pipe stress does not exceed the allowable stress. The pipeline modification was carried out by applied a gap of 3 mm in the X direction of the support at node 25 and eliminate the support at the node

  3. Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoder Jr, Graydon L [ORNL; Elkassabgi, Yousri M. [Texas A& M University, Kingsville; De Leon, Gerardo I. [Texas A& M University, Kingsville; Fetterly, Caitlin N. [Texas A& M University, Kingsville; Ramos, Jorge A. [Texas A& M University, Kingsville; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)

    2012-02-01

    Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental

  4. Update on Small Modular Reactors Dynamics System Modeling Tool -- Molten Salt Cooled Architecture

    Energy Technology Data Exchange (ETDEWEB)

    Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cetiner, Sacit M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Fugate, David L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Qualls, A L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Borum, Robert C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chaleff, Ethan S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rogerson, Doug W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Batteh, John J. [Modelon Corporation (Sweden); Tiller, Michael M. [Xogeny Corporation, Canton, MI (United States)

    2014-08-01

    The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the third year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled) concepts, including the use of multiple coupled reactors at a single site. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor SMR models, ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface (ICHMI) technical area, and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the Modular Dynamic SIMulation (MoDSIM) tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the program, (2) developing a library of baseline component modules that can be assembled into full plant models using existing geometry and thermal-hydraulic data, (3) defining modeling conventions for interconnecting component models, and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  6. Gas Cooled Fast Reactor Research and Development in the European Union

    Directory of Open Access Journals (Sweden)

    Richard Stainsby

    2009-01-01

    Full Text Available Gas-cooled fast reactor (GFR research is directed towards fulfilling the ambitious goals of Generation IV (Gen IV, that is, to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. The research is directed towards developing the GFR as an economic electricity generator, with good safety and sustainability characteristics. Fast reactors maximise the usefulness of uranium resources by breeding plutonium and can contribute to minimising both the quantity and radiotoxicity nuclear waste by actinide transmutation in a closed fuel cycle. Transmutation is particularly effective in the GFR core owing to its inherently hard neutron spectrum. Further, GFR is suitable for hydrogen production and process heat applications through its high core outlet temperature. As such GFR can inherit the non-electricity applications that will be developed for thermal high temperature reactors in a sustainable manner. The Euratom organisation provides a route by which researchers in all European states, and other non-European affiliates, can contribute to the Gen IV GFR system. This paper summarises the achievements of Euratom's research into the GFR system, starting with the 5th Framework programme (FP5 GCFR project in 2000, through FP6 (2005 to 2009 and looking ahead to the proposed activities within the 7th Framework Programme (FP7.

  7. Depletion Analysis of Modular High Temperature Gas-cooled Reactor Loaded with LEU/Thorium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sonat Sen; Gilles Youinou

    2013-02-01

    Thorium based fuel has been considered as an option to uranium-based fuel, based on considerations of resource utilization (Thorium is more widely available when compared to Uranium). The fertile isotope of Thorium (Th-232) can be converted to fissile isotope U-233 by neutron capture during the operation of a suitable nuclear reactor such as High Temperature Gas-cooled Reactor (HTGR). However, the fertile Thorium needs a fissile supporter to start and maintain the conversion process such as U-235 or Pu-239. This report presents the results of a study that analyzed the thorium utilization in a prismatic HTGR, namely Modular High Temperature Gas-Cooled Reactor (MHTGR) that was designed by General Atomics (GA). The collected for the modeling of this design come from Chapter 4 of MHTGR Preliminary Safety Information Document that GA sent to Department of Energy (DOE) on 1995. Both full core and unit cell models were used to perform this analysis using SCALE 6.1 and Serpent 1.1.18. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were set to match the spectral index between unit cell and full core domains. It was found that for the purposes of this study an adjusted unit cell model is adequate. Discharge isotopics and one-group cross-sections were delivered to the transmutation analysis team. This report provides documentation for these calculations

  8. Contributions to the neutronic analysis of a gas-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Martin-del-Campo, Cecilia, E-mail: cecilia.martin.del.campo@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Reyes-Ramirez, Ricardo, E-mail: ricarera@yahoo.com.mx [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Francois, Juan-Luis, E-mail: juan.luis.francois@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Reinking-Cejudo, Arturo G., E-mail: reinking@servidor.unam.mx [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico)

    2011-06-15

    Highlights: > Differences on reactivity with MCNPX and TRIPOLI-4 are negligible. > Fuel lattice and core criticality calculations were done. > A higher Doppler coefficient than coolant density coefficient. > Zirconium carbide is a better reflector than silicon carbide. > Adequate active height, radial size and reflector thickness were obtained. - Abstract: In this work the Monte Carlo codes MCNPX and TRIPOLI-4 were used to perform the criticality calculations of the fuel assembly and the core configuration of a gas-cooled fast reactor (GFR) concept, currently in development. The objective is to make contributions to the neutronic analysis of a gas-cooled fast reactor. In this study the fuel assembly is based on a hexagonal lattice of fuel-pins. The materials used are uranium and plutonium carbide as fuel, silicon carbide as cladding, and helium gas as coolant. Criticality calculations were done for a fuel assembly where the axial reflector thickness was varied in order to find the optimal thickness. In order to determine the best material to be used as a reflector, in the reactor core with neutrons of high energy spectrum, criticality calculations were done for three reflector materials: zirconium carbide, silicon carbide and natural uranium. It was found that the zirconium carbide provides the best neutron reflection. Criticality calculations using different active heights were done to determine the optimal height, and the reflector thickness was adjusted. Core criticality calculations were performed with different radius sizes to determine the active radial dimension of the core. A negative temperature coefficient of reactivity was verified for the fuel. The effect on reactivity produced by changes in the coolant density was also evaluated. We present the main neutronic characteristics of a preliminary fuel and core designs for the GFR concept. ENDF-VI cross-sections libraries were used in both the MCNPX and TRIPOLI-4 codes, and we verified that the obtained

  9. Final report-passive safety optimization in liquid sodium-cooled reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Cahalana, J. E.; Hahn, D.; Nuclear Engineering Division; Korea Atomic Energy Research Inst.

    2007-08-13

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4). Task 1--Computational Methods for Analysis of Passive Safety Design Features: An advanced three-dimensional subassembly thermal-hydraulic model was developed jointly and implemented in ANL and KAERI computer codes. The objective of the model development effort was to provide a high-accuracy capability to predict fuel, cladding, coolant, and structural temperatures in reactor fuel subassemblies, and thereby reduce the uncertainties associated with lower fidelity models previously used for safety and design analysis. The project included model formulation, implementation, and verification by application to available reactor tests performed at EBR-II. Task 2--Comparative Analysis and Evaluation of Innovative Design Features: Integrated safety assessments of innovative liquid metal reactor designs were performed to quantify the performance of inherent safety features. The objective of the analysis effort was to identify the potential safety margin enhancements possible in a sodium-cooled, metal-fueled reactor design by use of passive safety mechanisms to mitigate low-probability accident consequences. The project included baseline analyses using state-of-the-art computational models and advanced analyses using the new model developed in Task 1. Task 3--Safety

  10. Safe design of cooled tubular reactors for exothermic, multiple reactions; parallel reactions—II: The design and operation of an ethylene oxide reactor

    NARCIS (Netherlands)

    Westerterp, K.R.; Ptasiński, K.J.

    1984-01-01

    In part I a model and criteria have been developed for the safe design and operation of cooled tubular reactors for multiple reactions of the parallel type. In this Part II the model is extended to parallel reactions with an arbitrary stoichiometry. The results are applied to the industrial process

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

    Science.gov (United States)

    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.

  12. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  13. Toward a Mechanistic Source Term in Advanced Reactors: Characterization of Radionuclide Transport and Retention in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, Acacia J.; Bucknor, Matthew; Grabaskas, David

    2016-04-17

    A vital component of the U.S. reactor licensing process is an integrated safety analysis in which a source term representing the release of radionuclides during normal operation and accident sequences is analyzed. Historically, source term analyses have utilized bounding, deterministic assumptions regarding radionuclide release. However, advancements in technical capabilities and the knowledge state have enabled the development of more realistic and best-estimate retention and release models such that a mechanistic source term assessment can be expected to be a required component of future licensing of advanced reactors. Recently, as part of a Regulatory Technology Development Plan effort for sodium cooled fast reactors (SFRs), Argonne National Laboratory has investigated the current state of knowledge of potential source terms in an SFR via an extensive review of previous domestic experiments, accidents, and operation. As part of this work, the significant sources and transport processes of radionuclides in an SFR have been identified and characterized. This effort examines all stages of release and source term evolution, beginning with release from the fuel pin and ending with retention in containment. Radionuclide sources considered in this effort include releases originating both in-vessel (e.g. in-core fuel, primary sodium, cover gas cleanup system, etc.) and ex-vessel (e.g. spent fuel storage, handling, and movement). Releases resulting from a primary sodium fire are also considered as a potential source. For each release group, dominant transport phenomena are identified and qualitatively discussed. The key product of this effort was the development of concise, inclusive diagrams that illustrate the release and retention mechanisms at a high level, where unique schematics have been developed for in-vessel, ex-vessel and sodium fire releases. This review effort has also found that despite the substantial range of phenomena affecting radionuclide release, the

  14. Thermal Aspects of Using Alternative Nuclear Fuels in Supercritical Water-Cooled Reactors

    Science.gov (United States)

    Grande, Lisa Christine

    A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT)-type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties of reactor coolant, fuel sheath and fuel. Uniform and non-uniform AHFPs for average channel power were applied to a variety of alternative fuels (mixed oxide, thorium dioxide, uranium dicarbide, uranium nitride and uranium carbide) enclosed in an Inconel-600 43-element bundle. The results depict bulk-fluid, outer-sheath and fuel-centreline temperature profiles together with the Heat Transfer Coefficient (HTC) profiles along the heated length of fuel channel. The objective is to identify the best options in terms of fuel, sheath material and AHFPS in which the outer-sheath and fuel-centreline temperatures will be below the accepted temperature limits of 850°C and 1850°C respectively. The 43-element Inconel-600 fuel bundle is suitable for SCWR use as the sheath-temperature design limit of 850°C was maintained for all analyzed cases at average channel power. Thoria, UC2, UN and UC fuels for all AHFPs are acceptable since the maximum fuel-centreline temperature does not exceed the industry accepted limit of 1850°C. Conversely, the fuel-centreline temperature limit was exceeded for MOX at all AHFPs, and UO2 for both cosine and downstream-skewed cosine AHFPs. Therefore, fuel-bundle modifications are required for UO2 and MOX to be feasible nuclear fuels for SCWRs.

  15. A Cylindrical Shielding Design Concept for the Prototype Gen-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), a metal fueled, blanket-free, pool type SFR concept is adopted to acquire the inherent safety characteristics and high proliferation-resistance. In the pool type fast reactor, the intermediate heat exchangers (IHXs), which transfer heat from the primary sodium pool to a secondary sodium loop, are placed inside of the reactor vessel. Hence, secondary sodium passing the IHXs can be radioactivated by a {sup 23}Na(n,g){sup 24}Na reaction, and radioactivated secondary sodium causes a significant dose in the Steam Generator Building (SGB). Therefore, a typical core of a pool type fast reactor is usually surrounded by a massive quantity of shields. In addition, the blanket composed of depleted uranium plays a role as superior shielding material; a significant increase in shields is required in the blanket-free pool type SFR. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR. In a conventional shielding design, massive axial shields are required to prevent irradiation of secondary sodium passing IHXs and they should be replaced according to the subassembly replacement in spite of negligible depletion of the shielding material. The proposed shielding design concept minimizes the quantity of shields without their replacement. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR such as a PGSFR. The proposed design concept satisfied the dose limit in the steam generator building successfully without introducing a large quantity of B{sub 4}C shielding inside the subassembly.

  16. Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

    Science.gov (United States)

    Verma, V.; Barbot, L.; Filliatre, P.; Hellesen, C.; Jammes, C.; Svärd, S. Jacobsson

    2017-07-01

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

  17. Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

    Science.gov (United States)

    Garber, Anne E.; Dickens, Ricky E.

    2011-01-01

    The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

  18. Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hoon; Chang, Soon Heung; Kim, Soo Hyung; Kim, Kee Poong; Lee, Hyoung Wook; Jang, Kwang Keol; Jeong, Yong Hoon; Kim, Sang Jin; Lee, Seong Jin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Park, Jae Hong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2001-03-15

    In this work, assessment system for methodology for reactor pressure vessel integrity is developed. Assessment system is make up of severe accident assessment code which can calculate the conditions of plant and structural analysis code which can assess the integrity of reactor vessel using given plant conditions. An assessment of cavity flooding using containment spray system has been done. As a result, by the containment spray, cavity can be flooded successfully and CCI can be reduced. The technical backgrounds for external vessel cooling and corium cooling on the cavity are summarized and provided in this report.

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

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.; Baxi, C.B.; Reis, E.E. [General Atomics, San Diego, CA (United States); Cerbone, R.; Cheng, E.T. [TSI Research, Solana Beach, CA (United States)

    1998-03-01

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

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-15

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

  2. TRACG-CFD analysis of ESBWR reactor water cleanup shutdown cooling system mixing coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo, J. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Marquino, W.; Mistreanu, A.; Yang, J., E-mail: euqrop@hotmail.com [General Electric Hitachi Nuclear Energy, Wilmington, 28401 North Carolina (United States)

    2015-09-15

    The ESBWR is a 1520 nominal [M We] Generation III+ natural circulation boiling water reactor designed to high levels of safety utilizing features that have been successfully used before in operating BWRs, as well as standard features common to A BWR. In September of 2014, the US NRC has certified the ESBWR design for use in the USA. The RWCU/Sdc is an auxiliary system for the ESBWR nuclear island. Basic functions it performs include purifying the reactor coolant during normal operation and shutdown and providing shutdown cooling and cooldown to cold shutdown conditions. The performance of the RWCU system during shutdown cooling is directly related to the temperature of the water removed through the outlets, which is coupled with the vessel and F W temperatures through a thermal mixing coefficient. The complex three-dimensional (3-D) geometry of the BWR downcomer and lower plenum has a great impact on the flow mixing. Only a fine mesh technique like CFD can predict the 3-D temperature distribution in the RPV during shutdown and provide the RWCU/Sdc system inlet temperature. Plant shutdown is an unsteady event by nature and was modeled as a succession of CFD steady-state simulations. It is required to establish the mixing coefficient (which is a function of the heat balance and the core flow) during the operation of the RWCU system in the multiple shutdown cooling modes, and therefore a range of core flows needs to be estimated using quasi steady states obtained with TRACG. The lower end of that range is obtained from a system with minimal power decay heat and core flow; while the higher end corresponds to the power at the beginning of RWCU/Sdc operation when the cooldown is transferred to the RWCU/Sdc after the initial depressurization via the turbine bypass valves. Because the ESBWR RWCU/Sdc return and suction designs provide good mixing, the uniform mixing energy balance was found to be an adequate alternative for deriving the mixing coefficient. The CFD mass flow

  3. Cool dust heating and temperature mixing in nearby star-forming galaxies

    CERN Document Server

    Hunt, L K; Bianchi, S; Gordon, K D; Aniano, G; Calzetti, D; Dale, D A; Helou, G; Hinz, J L; Kennicutt, R C; Roussel, H; Wilson, C D; Bolatto, A; Boquien, M; Croxall, K V; Galametz, M; de Paz, A Gil; Koda, J; Munoz-Mateos, J C; Sandstrom, K M; Sauvage, M; Vigroux, L; Zibetti, S

    2014-01-01

    Physical conditions of the interstellar medium in galaxies are closely linked to the ambient radiation field and the heating of dust grains. In order to characterize dust properties in galaxies over a wide range of physical conditions, we present here the radial surface brightness profiles of the entire sample of 61 galaxies from Key Insights into Nearby Galaxies: Far-Infrared Survey with Herschel (KINGFISH). The main goal of our work is the characterization of the grain emissivities, dust temperatures, and interstellar radiation fields responsible for heating the dust. After fitting the dust and stellar radial profiles with exponential functions, we fit the far-infrared spectral energy distribution (SED) in each annular region with single-temperature modified black bodies using both variable (MBBV) and fixed (MBBF) emissivity indices beta, as well as with physically motivated dust models. Results show that while most SED parameters decrease with radius, the emissivity index beta also decreases with radius in...

  4. Temperature control of bench-scaled batch reactor equipped with a monofluid heating/cooling system

    Science.gov (United States)

    Teng, Hai-peng; Song, Yi-ming

    2014-04-01

    An advanced control concept, Predictive Functional Control (PFC), is applied for temperature control of a bench-scaled batch reactor equipped with monofluid heating/cooling system. First principles process models are developed. Based on achieved models, significant process variables, which are difficult or impossible to measure online, are estimated from easily measured variables, and cascade PFC control strategy has been projected and implemented in Matlab R14. The dynamics of individual subunits is explicitly taken into consideration by internal model in the control algorithms, and model uncertainty, various process disturbances are compensated by modification of internal model. The experimental results present an excellent capability of tracking the set point, and the success of PFC technique as a process control paradigm is illustratively demonstrated.

  5. Molecular Dynamics Simulations of Aqueous and Confined Systems Relevant to the Supercritical Water Cooled Nuclear Reactor

    Science.gov (United States)

    Kallikragas, Dimitrios Theofanis

    Supercritical water (SCW) is the intended heat transfer fluid and potential neutron moderator in the proposed GEN-IV Supercritical Water Cooled Reactor (SCWR). The oxidative environment poses challenges in choosing appropriate design materials, and the behaviour of SCW within crevices of the passivation layer is needed for developing a corrosion control strategy to minimize corrosion. Molecular Dynamics simulations have been employed to obtain diffusion coefficients, coordination number and surface density characteristics, of water and chloride in nanometer-spaced iron hydroxide surfaces. Diffusion models for hydrazine are evaluated along with hydration data. Results demonstrate that water is more likely to accumulate on the surface at low density conditions. The effect of confinement on the water structure diminishes as the gap size increases. The diffusion coefficient of chloride decreases with larger surface spacing. Clustering of water at the surface implies that the SCWR will be most susceptible to pitting corrosion and stress corrosion cracking.

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

    Science.gov (United States)

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

    2011-03-01

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

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

    Science.gov (United States)

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

    2013-09-03

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

  8. Fabrication of cermet bearings for the control system of a high temperature lithium cooled nuclear reactor

    Science.gov (United States)

    Yacobucci, H. G.; Heestand, R. L.; Kizer, D. E.

    1973-01-01

    The techniques used to fabricate cermet bearings for the fueled control drums of a liquid metal cooled reference-design reactor concept are presented. The bearings were designed for operation in lithium for as long as 5 years at temperatures to 1205 C. Two sets of bearings were fabricated from a hafnium carbide - 8-wt. % molybdenum - 2-wt. % niobium carbide cermet, and two sets were fabricated from a hafnium nitride - 10-wt. % tungsten cermet. Procedures were developed for synthesizing the material in high purity inert-atmosphere glove boxes to minimize oxygen content in order to enhance corrosion resistance. Techniques were developed for pressing cylindrical billets to conserve materials and to reduce machining requirements. Finishing was accomplished by a combination of diamond grinding, electrodischarge machining, and diamond lapping. Samples were characterized in respect to composition, impurity level, lattice parameter, microstructure and density.

  9. Using Wireless Sensor Networks to Achieve Intelligent Monitoring for High-Temperature Gas-Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Jianghai Li

    2017-01-01

    Full Text Available High-temperature gas-cooled reactors (HTGR can incorporate wireless sensor network (WSN technology to improve safety and economic competitiveness. WSN has great potential in monitoring the equipment and processes within nuclear power plants (NPPs. This technology not only reduces the cost of regular monitoring but also enables intelligent monitoring. In intelligent monitoring, large sets of heterogeneous data collected by the WSN can be used to optimize the operation and maintenance of the HTGR. In this paper, WSN-based intelligent monitoring schemes that are specific for applications of HTGR are proposed. Three major concerns regarding wireless technology in HTGR are addressed: wireless devices interference, cybersecurity of wireless networks, and wireless standards selected for wireless platform. To process nonlinear and non-Gaussian data obtained by WSN for fault diagnosis, novel algorithms combining Kernel Entropy Component Analysis (KECA and support vector machine (SVM are developed.

  10. NGNP: High Temperature Gas-Cooled Reactor Key Definitions, Plant Capabilities, and Assumptions

    Energy Technology Data Exchange (ETDEWEB)

    Phillip Mills

    2012-02-01

    This document is intended to provide a Next Generation Nuclear Plant (NGNP) Project tool in which to collect and identify key definitions, plant capabilities, and inputs and assumptions to be used in ongoing efforts related to the licensing and deployment of a high temperature gas-cooled reactor (HTGR). These definitions, capabilities, and assumptions are extracted from a number of sources, including NGNP Project documents such as licensing related white papers [References 1-11] and previously issued requirement documents [References 13-15]. Also included is information agreed upon by the NGNP Regulatory Affairs group's Licensing Working Group and Configuration Council. The NGNP Project approach to licensing an HTGR plant via a combined license (COL) is defined within the referenced white papers and reference [12], and is not duplicated here.

  11. Distribution and abundance of submerged aquatic macrophytes in a reactor cooling reservoir

    Energy Technology Data Exchange (ETDEWEB)

    Grace, J.B.

    1977-08-01

    Measurements of ash-free dry weight were used to characterize the effects of a heated effluent on submerged macrophytes in a reactor cooling reservoir. The species which were most abundant during the summers of 1974 and 1975 were Myriophyllum spicatum L. and Eleocharis acicularis (L.) R. and S. Examination of the vertical distribution of the shoot biomass of Myriophyllum revealed that plants in heated areas grew closer to the water surface than plants in unheated areas. The biomass of the second most abundant species, Eleocharis acicularis, was less at 0.5 m depths in heated areas (more than 5C/sup 0/ warmer than unheated areas) than at equal depths in unheated areas. Species diversity was greater at heated locations because of a greater equitability (i.e., evenness of distribution of biomass) among species.

  12. Simultaneous approach for simulation of a high-temperature gas-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    Yang CHEN; Jiang-hong YOU; Zhi-jiang SHAO; Ke-xin WANG; Ji-xin QIAN

    2011-01-01

    The simulation of a high-temperature gas-cooled reactor pebble-bed module (HTR-PM) plant is discussed.This lumped parameter model has the form of a set differential algebraic equations (DAEs) that include stiff equations to model point neutron kinetics.The nested approach is the most common method to solve DAE,but this approach is very expensive and time-consuming due to inner iterations.This paper deals with an alternative approach in which a simultaneous solution method is used.The DAEs are discretized over a time horizon using collocation on finite elements,and Radau collocation points are applied.The resulting nonlinear algebraic equations can be solved by existing solvers.The discrete algorithm is discussed in detail; both accuracy and stability issues are considered.Finally,the simulation results are presented to validate the efficiency and accuracy of the simultaneous approach that takes much less time than the nested one.

  13. Safety and licensing of MHTGR (Modular High Temperature Gas Cooled Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Silady, F.A.; Millunzi, A.C.; Kelley, A.P. Jr.; Cunliffe, J.

    1987-07-01

    The Modular High Temperature Gas Cooled Reactor (MHTGR) design meets stringent top-level regulatory and user safety requirements that require that the normal and off-normal operation of the plant not disturb the public's day-to-day activities. Quantitative, top-level regulatory criteria have been specified from US NRC and EPA sources to guide the design. The user/utility group has further specified that these criteria be met at the plant boundary. The focus of the safety approach has then been centered on retaining the radionuclide inventory within the fuel by removing core heat, controlling chemical attack, and by controlling heat generation. The MHTGR is shown to passively meet the stringent requirements with margin. No operator action is required and the plant is insensitive to operator error.

  14. High Temperature Gas-Cooled Reactor Projected Markets and Preliminary Economics

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2011-08-01

    This paper summarizes the potential market for process heat produced by a high temperature gas-cooled reactor (HTGR), the environmental benefits reduced CO2 emissions will have on these markets, and the typical economics of projects using these applications. It gives examples of HTGR technological applications to industrial processes in the typical co-generation supply of process heat and electricity, the conversion of coal to transportation fuels and chemical process feedstock, and the production of ammonia as a feedstock for the production of ammonia derivatives, including fertilizer. It also demonstrates how uncertainties in capital costs and financial factors affect the economics of HTGR technology by analyzing the use of HTGR technology in the application of HTGR and high temperature steam electrolysis processes to produce hydrogen.

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

    Science.gov (United States)

    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.

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

    Directory of Open Access Journals (Sweden)

    Aures Alexander

    2016-01-01

    Full Text Available 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.

  17. Development of the Sodium-cooled Fast Reactor R and D and Technology Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Uk; Won, Byung Chool; Kim, Young In; Hahn, Do Hee

    2008-01-15

    This study presents a R and D performance monitoring system that is applicable for managing the generation IV sodium-cooled fast reactor development. The prime goal of this system is to furnish project manager with reliable and accurate information of status of progress, performance and resource allocation, and attain traceability and visibility of project implementation for effective project management. In this study, the work breakdown structure, the related schedule and the expected outputs were established to derive the interfaces between projects and the above parameters was loaded PCs. The R and D performance monitoring system is composed of about 750 R and D activities within 'Development of Basic Key Technologies for Gen IV SFR' project in 2007. The Microsoft Project Professional software was used to monitor the progress, evaluate the results and analyze the resource distribution to activities.

  18. Use of Distribution Devices for Hydraulic Profiling of Coolant Flow in Core Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    A. A. Satin

    2014-01-01

    Full Text Available In setting up a reactor plant for the transportation-power module of the megawatt class an important task is to optimize the path of flow, i.e. providing moderate hydraulic resistance, uniform distribution of the coolant. Significant contribution to the hydraulic losses makes one selected design of the coolant supplies. It is, in particular, hemispherical or semi-elliptical shape of the supply reservoir, which is selected to reduce its mass, resulting in the formation of torusshaped vortex in the inlet manifold, that leads to uneven coolant velocity at the inlet into the core, the flow pulsations, hydraulic losses.To control the flow redistribution in the core according to the level of energy are used the switchgear - deflectors installed in a hemispherical reservoir supplying coolant to the fuel elements (FE of the core of gas-cooled reactor. This design solution has an effect on the structure of the flow, rate in the cooling duct, and the flow resistance of the collector.In this paper we present the results of experiments carried out on the gas dynamic model of coolant paths, deflectors, and core, comprising 55 fuel rod simulators. Numerical simulation of flow in two-parameter model, using the k-ε turbulence model, and the software package ANSYS CFX v14.0 is performed. The paper demonstrates that experimental results are in compliance with calculated ones.The results obtained suggest that the use of switchgear ensures a coolant flow balance directly at the core inlet, thereby providing temperature reduction of fuel rods with a uniform power release in the cross-section. Considered options to find constructive solutions for deflectors give an idea to solve the problem of reducing hydraulic losses in the coolant paths, to decrease pulsation components of flow in the core and length of initial section of flow stabilization.

  19. Validation of SCALE and the TRITON Depletion Sequence for Gas-Cooled Reactor Analysis

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, Mark D [ORNL; Pritchard, Megan L [ORNL

    2008-01-01

    The very-high-temperature reactor (VHTR) is an advanced reactor concept that uses graphite-moderated fuel and helium gas as a coolant. At present there are two primary VHTR reactor designs under consideration for development: in the pebble-bed reactor, a core is loaded with 'pebbles' consisting of 6 cm diameter spheres, while in a high-temperature gas-cooled reactor, fuel rods are placed within prismatic graphite blocks. In both systems, fuel elements (spheres or rods) are comprised of tristructural-isotropic (TRISO) fuel particles. The TRISO particles are either dispersed in the matrix of a graphite pebble for the pebble-bed design or molded into compacts/rods that are then inserted into the hexagonal graphite blocks for the prismatic concept. Two levels of heterogeneity exist in such fuel designs: (1) microspheres of TRISO particles dispersed in a graphite matrix of a cylindrical or spherical shape, and (2) neutron interactions at the rod-to-rod or sphere-to-sphere level. Such double heterogeneity (DH) provides a challenge to multigroup cross-section processing methods, which must treat each level of heterogeneity separately. A new capability to model doubly heterogeneous systems was added to the SCALE system in the release of Version 5.1. It was included in the control sequences CSAS and CSAS6, which use the Monte Carlo codes KENO V.a and KENO-VI, respectively, for three-dimensional neutron transport analyses and in the TRITON sequence, which uses the two-dimensional lattice physics code NEWT along with both versions of KENO for transport and depletion analyses. However, the SCALE 5.1 version of TRITON did not support the use of the DH approach for depletion. This deficiency has been addressed, and DH depletion will be available as an option in the upcoming release of SCALE 6. At present Oak Ridge National Laboratory (ORNL) staff are developing a set of calculations that may be used to validate SCALE for DH calculations. This paper discusses the

  20. Measurement of Turbulent Flow Phenomena for the Lower Plenum of a Prismatic Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hugh M. McIlroy Jr.; Donald M. McEligot; Robert J. Pink; Keith G. Condie; Glenn E. McCreery

    2007-09-01

    Mean velocity field and turbulence data are presented for flow phenomena in a lower plenum of a typical prismatic gas-cooled reactor (GCR), such as in a Very High Temperature Reactor (VHTR) concept. In preparation for design, safety analyses and licensing, research has begun on readying the computational tools that will be needed to predict the thermal-hydraulics behavior of the reactor design. Fluid dynamics experiments have been designed and built to develop benchmark databases for the assessment of computational fluid dynamics (CFD) codes and their turbulence models for a typical VHTR plenum geometry in the limiting case of negligible buoyancy and constant fluid properties. This experiment has been proposed as a “Standard Problem” for assessing advanced reactor (CFD) analysis tools. Present results concentrate on the region of the plenum near its far reflector wall (away from the outlet duct). The flow in the lower plenum can locally be considered as multiple jets into a confined cross flow - with obstructions. A model of the lower plenum has been fabricated and scaled to the geometric dimensions of the Next Generation Nuclear Plant (NGNP) Point Design. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to induce flow features somewhat comparable to those expected from the staggered parallel rows of posts in the reactor design. Posts, side walls and end walls are fabricated from clear, fused quartz to match the refractive-index of the working fluid so that optical techniques may be employed for the measurements. The experiments were conducted in the Matched-Index-of-Refraction (MIR) Facility at the Idaho National Laboratory (INL). The benefit of the MIR technique is that it permits optical measurements to determine complex flow characteristics in passages and around objects to be obtained without locating a disturbing transducer in the flow field and without distortion of the optical paths. The

  1. Army Gas-Cooled Reactor Systems Program. Operation of ML-1 reactor skid in GCRE: safety evaluation report

    Energy Technology Data Exchange (ETDEWEB)

    None

    1964-10-01

    The operation of the ML-1 reactor skid in the modified GCRE facility, utilizing the GCRE reactor coolant circulating and heat removal systems, is described. An evaluation of the safety considerations associated with this mode of operation indicates that the consequences of the maximum credible accident are less severe than those previously approved for operation of the ML-1 reactor at the ML-1 test site or for operation of the GCRE-I reactor in the GCRE facility.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

    This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigation of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, the 400 MWt Fast Flux Test Facility (FFTF) achieved a peak neutron fast flux of 7.2 x 10{sup

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

    Energy Technology Data Exchange (ETDEWEB)

    Ponciroli, R.; Passerini, S.; Vilim, R. B.

    2016-04-17

    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.

  4. Study of guided wave transmission through complex junction in sodium cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Elie, Q.; Le Bourdais, F.; Jezzine, K.; Baronian, V. [Non Destructive Testing Department at the French Atomic Energy Commission (CEA), Saclay, 91191 Gif sur Yvette CEDEX, (France)

    2015-07-01

    Ultrasonic guided wave techniques are seen as suitable candidates for the inspection of welded structures within sodium cooled fast reactors (SFR), as the long range propagation of guided waves without amplitude attenuation can overcome the accessibility problem due to the liquid sodium. In the context of the development of the Advanced Sodium Test Reactor for Industrial Demonstration (ASTRID), the French Atomic Commission (CEA) investigates non-destructive testing techniques based on guided wave propagation. In this work, guided wave NDT methods are applied to control the integrity of welds located in a junction-type structure welded to the main vessel. The method presented in this paper is based on the analysis of scattering matrices peculiar to each expected defect, and takes advantage of the multi-modal and dispersive characteristics of guided wave generation. In a simulation study, an algorithm developed using the CIVA software is presented. It permits selecting appropriate incident modes to optimize detection and identification of expected flawed configurations. In the second part of this paper, experimental results corresponding to a first validation step of the simulation results are presented. The goal of the experiments is to estimate the effectiveness of the incident mode selection in plates. The results show good agreement between experience and simulation. (authors)

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

    Directory of Open Access Journals (Sweden)

    T. R. Allen

    2007-01-01

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

  6. Procedure of Active Residual Heat Removal after Emergency Shutdown of High-Temperature-Gas-Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Xingtuan Yang

    2014-01-01

    Full Text Available After emergency shutdown of high-temperature-gas-cooled reactor, the residual heat of the reactor core should be removed. As the natural circulation process spends too long period of time to be utilized, an active residual heat removal procedure is needed, which makes use of steam generator and start-up loop. During this procedure, the structure of steam generator may suffer cold/heat shock because of the sudden load of coolant or hot helium at the first few minutes. Transient analysis was carried out based on a one-dimensional mathematical model for steam generator and steam pipe of start-up loop to achieve safety and reliability. The results show that steam generator should be discharged and precooled; otherwise, boiling will arise and introduce a cold shock to the boiling tubes and tube sheet when coolant began to circulate prior to the helium. Additionally, in avoiding heat shock caused by the sudden load of helium, the helium circulation should be restricted to start with an extreme low flow rate; meanwhile, the coolant of steam generator (water should have flow rate as large as possible. Finally, a four-step procedure with precooling process of steam generator was recommended; sensitive study for the main parameters was conducted.

  7. High Temperature Fission Chamber for He- and FLiBe-cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bell, Zane W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giuliano, Dominic R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holcomb, David Eugene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lance, Michael J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, Roger G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Warmack, Robert J. Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Mark J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    We have evaluated candidate technologies for in-core fission chambers for high-temperature reactors to monitor power level via measurements of neutron flux from start-up through full power at up to 800°C. This research is important because there are no commercially available instruments capable of operating above 550 °C. Component materials and processes were investigated for fission chambers suitable for operation at 800 °C in reactors cooled by molten fluoride salt (FLiBe) or flowing He, with an emphasis placed on sensitivity (≥ 1 cps/nv), service lifetime (2 years at full power), and resistance to direct immersion in FLiBe. The latter gives the instrument the ability to survive accidents involving breach of a thimble. The device is envisioned to be a two-gap, three-electrode instrument constructed from concentric nickel-plated alumina cylinders and using a noble gas–nitrogen fill-gas. We report the results of measurements and calculations of the response of fill gasses, impurity migration in nickel alloy, brazing of the alumina insulator, and thermodynamic calculations.

  8. Coupling of Modular High-Temperature Gas-Cooled Reactor with Supercritical Rankine Cycle

    Directory of Open Access Journals (Sweden)

    Shutang Zhu

    2008-01-01

    Full Text Available This paper presents investigations on the possible combination of modular high-temperature gas-cooled reactor (MHTGR technology with the supercritical (SC steam turbine technology and the prospective deployments of the MHTGR SC power plant. Energy conversion efficiency of steam turbine cycle can be improved by increasing the main steam pressure and temperature. Investigations on SC water reactor (SCWR reveal that the development of SCWR power plants still needs further research and development. The MHTGR SC plant coupling the existing technologies of current MHTGR module design with operation experiences of SC FPP will achieve high cycle efficiency in addition to its inherent safety. The standard once-reheat SC steam turbine cycle and the once-reheat steam cycle with life-steam have been studied and corresponding parameters were computed. Efficiencies of thermodynamic processes of MHTGR SC plants were analyzed, while comparisons were made between an MHTGR SC plant and a designed advanced passive PWR - AP1000. It was shown that the net plant efficiency of an MHTGR SC plant can reach 45% or above, 30% higher than that of AP1000 (35% net efficiency. Furthermore, an MHTGR SC plant has higher environmental competitiveness without emission of greenhouse gases and other pollutants.

  9. Preliminary Study on the High Efficiency Supercritical Pressure Water-Cooled Reactor for Electricity Generation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yoon Yeong; Park, Jong Kyun; Cho, Bong Hyun and others

    2006-01-15

    This research has been performed to introduce a concept of supercritical pressure water cooled reactor(SCWR) in Korea The area of research includes core conceptual design, evaluation of candidate fuel, fluid systems conceptual design with mechanical consideration, preparation of safety analysis code, and construction of supercritical pressure heat transfer test facility, SPHINX, and preliminary test. As a result of the research, a set of tools for the reactor core design has been developed and the conceptual core design with solid moderator was proposed. The direct thermodynamic cycle has been studied to find a optimum design. The safety analysis code has also been adapted to supercritical pressure condition. A supercritical pressure CO2 heat transfer test facility has been constructed and preliminary test proved the facility works as expected. The result of this project will be good basis for the participation in the international collaboration under GIF GEN-IV program and next 5-year mid and long term nuclear research program of MOST. The heat transfer test loop, SPHINX, completed as a result of this project may be used for the power cycle study as well as further heat transfer study for the various geometries.

  10. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  11. DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-TEMPERATURE GAS-COOLED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James; Bayless, Paul; Strydom, Gerhard; Kumar, Akansha; Gougar, Hans

    2016-11-01

    Uncertainty and sensitivity analysis is an indispensable element of any substantial attempt in reactor simulation validation. The quantification of uncertainties in nuclear engineering has grown more important and the IAEA Coordinated Research Program (CRP) on High-Temperature Gas Cooled Reactor (HTGR) initiated in 2012 aims to investigate the various uncertainty quantification methodologies for this type of reactors. The first phase of the CRP is dedicated to the estimation of cell and lattice model uncertainties due to the neutron cross sections co-variances. Phase II is oriented towards the investigation of propagated uncertainties from the lattice to the coupled neutronics/thermal hydraulics core calculations. Nominal results for the prismatic single block (Ex.I-2a) and super cell models (Ex.I-2c) have been obtained using the SCALE 6.1.3 two-dimensional lattice code NEWT coupled to the TRITON sequence for cross section generation. In this work, the TRITON/NEWT-flux-weighted cross sections obtained for Ex.I-2a and various models of Ex.I-2c is utilized to perform a sensitivity analysis of the MHTGR-350 core power densities and eigenvalues. The core solutions are obtained with the INL coupled code PHISICS/RELAP5-3D, utilizing a fixed-temperature feedback for Ex. II-1a.. It is observed that the core power density does not vary significantly in shape, but the magnitude of these variations increases as the moderator-to-fuel ratio increases in the super cell lattice models.

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

    Directory of Open Access Journals (Sweden)

    Zhu Guifeng

    2016-01-01

    Full Text Available Sustainability of thorium fuel in a Pebble-Bed Fluoride salt-cooled High temperature Reactor (PB-FHR is investigated to find the feasible region of high discharge burnup and negative Flibe (2LiF-BeF2 salt Temperature Reactivity Coefficient (TRC. Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tristructural-isotropic (TRISO coated particle system for increasing fuel loading and decreasing excessive moderation. To analyze the neutronic characteristics, an equilibrium calculation method of thorium fuel self-sustainability is developed. We have compared two refueling schemes (mixing flow pattern and directional flow pattern and two kinds of reflector materials (SiC and graphite. This method found that the feasible region of breeding and negative Flibe TRC is between 20 vol% and 62 vol% fuel loading in the fuel. A discharge burnup could be achieved up to about 200 MWd/kgHM. The case with directional flow pattern and SiC reflector showed superior burnup characteristics but the worst radial power peak factor, while the case with mixing flow pattern and SiC reflector, which was the best tradeoff between discharge burnup and radial power peak factor, could provide burnup of 140 MWd/kgHM and about 1.4 radial power peak factor with 50 vol% dispersion fuel. In addition, Flibe salt displays good neutron properties as a coolant of quasi-fast reactors due to the strong 9Be(n,2n reaction and low neutron absorption of 6Li (even at 1000 ppm in fast spectrum. Preliminary thermal hydraulic calculation shows good safety margin. The greatest challenge of this reactor may be the decades irradiation time of the pebble fuel.

  13. A Conceptual Study on a Supercritical CO2-cooled KAIST Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yu, HwanYeal; Hartanto, Donny; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Small Modular Reactors (SMRs) are nuclear reactors that are completely built in a factory and shipped to the designated site for installation. As such, the SMR is especially advantageous as a flexible and cost-effective energy source for remote and isolated areas. Furthermore, the concept requires a relatively low capital cost, which makes it attractive for developing countries with limited electricity grid. In addition, the SMR concepts also generate more interest after the Fukushima accident since it can easily be designed with a passive decay heat removal system. One of the major advantages of a water-cooled SMR is its relatively small core size. Nonetheless, in spite of its small core size, the volume and area required for its steam-cycle power conversion unit is still significant. In this study, neutronics feasibility of a fully compact and transportable KAIST micro-modular reactor (MMR) was demonstrated. Rated thermal power of the core was 36.2 MWth with total weight of about 39.6 tons. The core was loaded with 15.5 w/o uranium mono-nitride U15N fuels in order to achieve a targeted lifetime of 20 EFPYs. To achieve targeted lifetime, initial excess reactivity of the core should be quite high, around 4,707 pcm. To reduce the high excess reactivity to about 2,500 pcm, a replaceable burnable absorber was utilized in the design. As a result, the MMR has a 20-year lifetime with a relatively small burnup reactivity swing. Several important safety parameters of the KAIST MMR core were also determined in this study. The Doppler reactivity coefficients and CVRs were demonstrated to negative. Worth of the primary control drums and secondary control rod were much higher than initial excess reactivity.

  14. An experimental study of the selective oxidation of ethene in a wall cooled tubular packed bed reactor

    NARCIS (Netherlands)

    Borman, P.C.; Westerterp, K.R.

    1992-01-01

    The selective oxidation of ethene over a silver on ¿-alumina catalyst was studied in a wall cooled tubular reactor. Temperatures were measured inside the bed at different axial and radical positions as well as the overall conversion and selectivity. Locally measured temperatures vary after repacking

  15. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-3: High Temperature Gas Cooled Reactor Thermal-Hydraulics.

    Science.gov (United States)

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical high temperature gas-cooled reactor (HTGR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module…

  16. Time dependence of corrosion in steels for use in lead-alloy cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Machut, McLean [Department of Nuclear Engineering and Engineering Physics, University of Wisconsin-Madison, 1500 Engineering Dr., Madison, WI 53706 (United States)], E-mail: mtmachut@wisc.edu; Sridharan, Kumar [Department of Nuclear Engineering and Engineering Physics, University of Wisconsin-Madison, 1500 Engineering Dr., Madison, WI 53706 (United States); Li Ning [Materials Physics and Application Division, AFCI, Los Alamos National Laboratory, NM (United States); Ukai, Shigeharu [Division of Materials Science and Engineering, Hokaido University (Japan); Allen, Todd [Department of Nuclear Engineering and Engineering Physics, University of Wisconsin-Madison, 1500 Engineering Dr., Madison, WI 53706 (United States)

    2007-09-15

    Stability of the protective oxide layer is critical for the long-term performance of cladding and structural components in lead-alloy cooled nuclear systems. Measurements have shown that removal of the outer magnetite layer is a significant effect at higher temperatures in flowing lead-bismuth. Developing a predictive capability for oxide thickness and material removal is therefore needed. A model for the corrosion of steels in liquid lead-alloys has been employed to assist in materials development for application in the Generation IV Lead-cooled Fast Reactor (LFR). Data from corrosion tests of steels in Los Alamos National Laboratory's DELTA Loop is used to benchmark the model and to obtain predictions of long-term material's corrosion performance. The model is based on modifications of Wagner's diffusion based oxidation theory and Tedmon's equation for high-temperature oxidation with scale removal. Theoretically and experimentally obtained values for parabolic oxide growth rate, mass transfer corrosion rate, and long-term material thinning rates are presented and compared to the literature.

  17. Current design efforts for the gas-cooled fast reactor (GFR)

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, K.D. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, Idaho 83415-3850 (United States)]. e-mail: Kevan.Weaver@inl.gov

    2005-07-01

    Current research and development on the Gas-Cooled Fast Reactor (GCFR) 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 (AFC I) 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 GCFR: a helium-cooled, direct Brayton cycle power conversion system that will operate with an 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 GCFR. These are EURATOM (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, EURATOM (including the United Kingdom), France, Japan, and Switzerland have active research activities with respect to the GCFR. The research includes GCFR design and safety, and fuels/in-core materials/fuel cycle projects. This paper outlines the current design status of the GCFR, and includes work done in the areas mentioned above. (Author)

  18. High-Temperature Gas-Cooled Reactor Technology Development Program: Annual progress report for period ending December 31, 1987

    Energy Technology Data Exchange (ETDEWEB)

    Jones, J.E.,Jr.; Kasten, P.R.; Rittenhouse, P.L.; Sanders, J.P.

    1989-03-01

    The High-Temperature Gas-Cooled Reactor (HTGR) Program being carried out under the US Department of Energy (DOE) continues to emphasize the development of modular high-temperature gas-cooled reactors (MHTGRs) possessing a high degree of inherent safety. The emphasis at this time is to develop the preliminary design of the reference MHTGR and to develop the associated technology base and licensing infrastructure in support of future reactor deployment. A longer-term objective is to realize the full high-temperature potential of HTGRs in gas turbine and high-temperature, process-heat applications. This document summarizes the activities of the HTGR Technology Development Program for the period ending December 31, 1987.

  19. Analysis of the horizontal flow in the advanced gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Duan, Y. [Department of Mechanical Engineering, University of Sheffield, Sheffield S1 3JD (United Kingdom); He, S., E-mail: s.he@sheffield.ac.uk [Department of Mechanical Engineering, University of Sheffield, Sheffield S1 3JD (United Kingdom); Ganesan, P. [Department of Mechanical Engineering, University of Malaya, Kuala Lumpur 50603 (Malaysia); Gotts, J. [EDF Energy, Barnwood, Gloucester GL4 3RS (United Kingdom)

    2014-06-01

    Highlights: • CFD is used to assess the effect of horizontal flows in AGRs. • The horizontal flows can reduce the graphite brick temperature significantly. • Such effects are not taken into consideration in current engineering calculations. • There might be flow instabilities when the fuel channel flow is very low but horizontal flows reduce its possibility. - Abstract: The purpose of the paper is to report a computational investigation of horizontal flows in the UK advanced-gas-cooled reactor (AGR) by using computational fluid dynamics with ANSYS FLUENT. The study is relevant to practical issues encountered in some AGR stations currently in operation in the UK. It is carried out using a comparative approach based on the results of two contrasting models: one simulating the full effect of the cross flow, the other simulating the simplified approach currently employed by the industry which neglects the momentum of the horizontal cross flow. The study reveals that the horizontal cross flow plays a significant role in the cooling of the moderator brick, while the axial variation of the brick geometry also significantly changes the distribution of the temperature within the brick. It is also found that under some circumstances the so-called horizontal inter-brick leakage (HIBL) flow could influence the cooling performance in the narrow gaps, resulting in a local hot spot. Furthermore, there may be flow instabilities in the flows in AGR fuel channels due to the interactions between the flow in the main arrowhead flow passages and that in some narrow passages connected to it, but the influence on the brick temperature is negligible. Horizontal cross flow has an effect of reducing such instabilities.

  20. Secondary Cooling Water Quality Management for Multi Purpose Reactor 30 MW GA Siwabessy Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Sunaryo, Geni Rina, E-mail: genirina@batan.go.i [Center for Reactor Technology and Nuclear Safety (PTRKN-BATAN), Bldg. 80, Puspiptek Area, Serpong, Tangerang 15310 (Indonesia)

    2011-07-01

    Indonesia Multi Purpose Research Reactor (MPR) G.A. Siwabessy 30 MW will be 25 years old in 2011. Series of Non Destructive Test (NDT) were done to understand the current condition such as Eddy Current test for Heat Exchangers, water immersed camera for understanding the tank liner condition, ultrasonic for secondary piping etc. Some deteorization was observed because of ageing and some changing was done. One of them is changing some part of secondary pipe lines because of leaking, with the local ones. For having another 25 years operation life, a proper water quality for secondary cooling water is needed towards corrosion prevention. The main objectives of this experiment is to understand the current water quality of secondary cooling water of RSG-GAS from the aspect of corrosion induced by chemicals and bacteria, and establish procedure for managing the secondary cooling water quality. Methodologies applied are surveillance corrosion by immersing coupon into water observed and followed by visual analyses, corrosion rate determination by electrochemical method with various chemical conditions and total bacteria determination by using test kit. The results show visually that the crevice, galvanic and homogeny corrosion with the current water quality easily be observed for carbon steel represented secondary pipelines at the condition of none oxy bio agent addition. This corrosion is being suppressed by adding the oxy bio agent. The orientation of coupon, vertically and horizontally, gives slightly different effect. The closely corrosion rate was obtained by separately experiment, electrochemical, at the concentration of inhibitor 100ppm is 0.13 {+-} 0.02, which is lower than in the raw water of 0.20 {+-} 0.01 mpy. The total bacteria detected is around 10{sup 7} cfu/ml at none reactor operation and without any anti bacteria added. The oxi bio agent chemical addition suppresses the numbers becomes 10{sup 3} cfu/ml. The SRB bacteria is detected as >10{sup 6} cfu/ml at

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-02-01

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

  2. Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

    Science.gov (United States)

    Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

    2017-01-01

    In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

  3. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, September 23, 1976--December 31, 1976

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    This report presents the results of work performed from September 23, 1976 through December 31, 1976 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed.

  4. Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1979-September 30, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1980-03-07

    The results of work performed from July 1, 1979 through September 30, 1979 on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented. In addition, the progress in the screening test program is described.

  5. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, January 1, 1980-March 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-25

    Results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Included are the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described, including screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, and 950/sup 0/C.

  6. Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, April 1, 1979-June 30, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-25

    The results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented. In addition, the progress in the screening test program is described.

  7. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, July 1--September 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-11-24

    Results of work performed from July 1, 1978 through September 30, 1978 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. Candidate alloys were evaluated for Very High Temperature Reactor Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (Helium containing small amounts of various other gases), the high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. The activities associated with the characterization of the materials for the screening test program are reported, i.e., test specimen preparation, information from the materials characterization tests performed by General Electric, and the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented.

  8. Experimental study on the heat transfer characteristics of a nuclear reactor containment wall cooled by gravitationally falling water

    Science.gov (United States)

    Pasek, Ari D.; Umar, Efrison; Suwono, Aryadi; Manalu, Reinhard E. E.

    2012-06-01

    Gravitationally falling water cooling is one of mechanism utilized by a modern nuclear Pressurized Water Reactor (PWR) for its Passive Containment Cooling System (PCCS). Since the cooling is closely related to the safety, water film cooling characteristics of the PCCS should be studied. This paper deals with the experimental study of laminar water film cooling on the containment model wall. The influences of water mass flow rate and wall heat rate on the heat transfer characteristic were studied. This research was started with design and assembly of a containment model equipped with the water cooling system, and calibration of all measurement devices. The containment model is a scaled down model of AP 1000 reactor. Below the containment steam is generated using electrical heaters. The steam heated the containment wall, and then the temperatures of the wall in several positions were measure transiently using thermocouples and data acquisition. The containment was then cooled by falling water sprayed from the top of the containment. The experiments were done for various wall heat rate and cooling water flow rate. The objective of the research is to find the temperature profile along the wall before and after the water cooling applied, prediction of the water film characteristic such as means velocity, thickness and their influence to the heat transfer coefficient. The result of the experiments shows that the wall temperatures significantly drop after being sprayed with water. The thickness of water film increases with increasing water flow rate and remained constant with increasing wall heat rate. The heat transfer coefficient decreases as film mass flow rate increase due to the increases of the film thickness which causes the increasing of the thermal resistance. The heat transfer coefficient increases slightly as the wall heat rate increases. The experimental results were then compared with previous theoretical studied.

  9. Conceptual design for accelerator-driven sodium-cooled sub-critical transmutation reactors using scale laws

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwang Gu; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    The feasibility study on conceptual design methodology for accelerator-driven sodium-cooled sub-critical transmutation reactors has been conducted to optimize the design parameters from the scale laws and validates the reactor performance with the integrated code system. A 1000 MWth sodium-cooled sub-critical transmutation reactor has been scaled and verified through the methodology in this paper, which is referred to Advanced Liquid Metal Reactor (ALMR). A Pb-Bi target material and a partitioned fuel are the liquid phases, and they are cooled by the circulation of secondary Pb-Bi coolant and by primary sodium coolant, respectively. Overall key design parameters are generated from the scale laws and they are improved and validated by the integrated code system. Integrated Code System (ICS) consists of LAHET, HMCNP, ORIGEN2, and COMMIX codes and some files. Through ICS the target region, the core region, and thermal-hydraulic related regions are analyzed once-through Results of conceptual design are attached in this paper. 5 refs., 4 figs., 1 tab. (Author)

  10. TRISO-Coated Fuel Processing to Support High Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Del Cul, G.D.

    2002-10-01

    The initial objective of the work described herein was to identify potential methods and technologies needed to disassemble and dissolve graphite-encapsulated, ceramic-coated gas-cooled-reactor spent fuels so that the oxide fuel components can be separated by means of chemical processing. The purpose of this processing is to recover (1) unburned fuel for recycle, (2) long-lived actinides and fission products for transmutation, and (3) other fission products for disposal in acceptable waste forms. Follow-on objectives were to identify and select the most promising candidate flow sheets for experimental evaluation and demonstration and to address the needs to reduce technical risks of the selected technologies. High-temperature gas-cooled reactors (HTGRs) may be deployed in the next -20 years to (1) enable the use of highly efficient gas turbines for producing electricity and (2) provide high-temperature process heat for use in chemical processes, such as the production of hydrogen for use as clean-burning transportation fuel. Also, HTGR fuels are capable of significantly higher burn-up than light-water-reactor (LWR) fuels or fast-reactor (FR) fuels; thus, the HTGR fuels can be used efficiently for transmutation of fissile materials and long-lived actinides and fission products, thereby reducing the inventory of such hazardous and proliferation-prone materials. The ''deep-burn'' concept, described in this report, is an example of this capability. Processing of spent graphite-encapsulated, ceramic-coated fuels presents challenges different from those of processing spent LWR fuels. LWR fuels are processed commercially in Europe and Japan; however, similar infrastructure is not available for processing of the HTGR fuels. Laboratory studies on the processing of HTGR fuels were performed in the United States in the 1960s and 1970s, but no engineering-scale processes were demonstrated. Currently, new regulations concerning emissions will impact the

  11. Analysis of Fluid Flow and Heat Transfer Model for the Pebble Bed High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    S. Yamoah

    2012-06-01

    Full Text Available The pebble bed type high temperature gas cooled nuclear reactor is a promising option for next generation reactor technology and has the potential to provide high efficiency and cost effective electricity generation. The reactor unit heat transfer poses a challenge due to the complexity associated with the thermalflow design. Therefore to reliably simulate the flow and heat transport of the pebble bed modular reactor necessitates a heat transfer model that deals with radiation as well as thermal convection and conduction. In this study, a model with the capability to simulate fluid flow and heat transfer in the pebble bed modular reactor core has been developed. The developed model was implemented on a personal computer using FORTRAN 95 programming language. Several important fluid flow and heat transfer parameters have been examined: including the pressure drop over the reactor core, the heat transfer coefficient, the Nusselt number and the effective thermal conductivity of the fuel pebbles. Results obtained from the simulation experiments show a uniform pressure in the radial direction for a core to fuel element diameter (D/d ratio>20 and the heat transfer coefficient increases with increasing temperature and coolant mass flow rate. The model can adequately account for the flow and heat transfer phenomenon and the loss of pressure through friction in the pebble bed type high temperature nuclear reactor.

  12. ALMA observations of cool dust in a low-metallicity starburst, SBS0335-052

    CERN Document Server

    Hunt, L K; Casasola, V; Garcia-Burillo, S; Combes, F; Nikutta, R; Caselli, P; Henkel, C; Maiolino, R; Menten, K M; Sauvage, M; Weiss, A

    2013-01-01

    We present Atacama Large Millimeter/submillimeter Array (ALMA) Cycle 0 Band 7 observations of an extremely metal-poor dwarf starburst galaxy in the Local Universe, SBS0335-052 (12+log(O/H)~7.2). With these observations, dust is detected at 870micron (ALMA Band 7), but 87% of the flux in this band is due to free-free emission from the starburst. We have compiled a spectral energy distribution (SED) of SBS0335-052 that spans almost 6 orders of magnitude in wavelength and fit it with a spherical dust shell heated by a single-age stellar population; the best-fit model gives a dust mass of (3.8+/-0.6)x10^4 Msun. We have also constructed a SED including Herschel archival data for IZw18, another low-metallicity dwarf starburst (12+log(O/H)=7.17), and fit it with a similar model to obtain a dust mass of (3.4+/-1.0)x10^2 Msun. Compared with their atomic gas mass, the dust mass of SBS0335-052 far exceeds the prediction of a linear trend of dust-to-gas mass ratio with metallicity, while IZw18 falls far below. We use gas...

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Approaches to experimental validation of high-temperature gas-cooled reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Belov, S.E. [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Borovkov, M.N., E-mail: borovkov@okbm.nnov.ru [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Golovko, V.F.; Dmitrieva, I.V.; Drumov, I.V.; Znamensky, D.S.; Kodochigov, N.G. [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Baxi, C.B.; Shenoy, A.; Telengator, A. [General Atomics, 3550 General Atomics Court, CA (United States); Razvi, J., E-mail: Junaid.Razvi@ga.com [General Atomics, 3550 General Atomics Court, CA (United States)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer Computational and experimental investigations of thermal and hydrodynamic characteristics for the equipment. Black-Right-Pointing-Pointer Vibroacoustic investigations. Black-Right-Pointing-Pointer Studies of the electromagnetic suspension system on GT-MHR turbo machine rotor models. Black-Right-Pointing-Pointer Experimental investigations of the catcher bearings design. - Abstract: The special feature of high-temperature gas-cooled reactors (HTGRs) is stressed operating conditions for equipment due to high temperature of the primary circuit helium, up to 950 Degree-Sign C, as well as acoustic and hydrodynamic loads upon the gas path elements. Therefore, great significance is given to reproduction of real operation conditions in tests. Experimental investigation of full-size nuclear power plant (NPP) primary circuit components is not practically feasible because costly test facilities will have to be developed for the power of up to hundreds of megawatts. Under such conditions, the only possible process to validate designs under development is representative tests of smaller scale models and fragmentary models. At the same time, in order to take in to validated account the effect of various physical factors, it is necessary to ensure reproduction of both individual processes and integrated tests incorporating needed integrated investigations. Presented are approaches to experimental validation of thermohydraulic and vibroacoustic characteristics for main equipment components and primary circuit path elements under standard loading conditions, which take account of their operation in the HTGR. Within the framework of the of modular helium reactor project, including a turbo machine in the primary circuit, a new and difficult problem is creation of multiple-bearing flexible vertical rotor. Presented are approaches to analytical and experimental validation of the rotor electromagnetic bearings, catcher bearings, flexible rotor

  15. Pre-Conceptual Design of a Fluoride-Salt-Cooled Small Modular Advanced High Temperature Reactor (SmAHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Greene, Sherrell R [ORNL; Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Carbajo, Juan J [ORNL; Ilas, Dan [ORNL; Cisneros, Anselmo T [ORNL; Varma, Venugopal Koikal [ORNL; Corwin, William R [ORNL; Wilson, Dane F [ORNL; Yoder Jr, Graydon L [ORNL; Qualls, A L [ORNL; Peretz, Fred J [ORNL; Flanagan, George F [ORNL; Clayton, Dwight A [ORNL; Bradley, Eric Craig [ORNL; Bell, Gary L [ORNL; Hunn, John D [ORNL; Pappano, Peter J [ORNL; Cetiner, Sacit M [ORNL

    2011-02-01

    This document presents the results of a study conducted at Oak Ridge National Laboratory during 2010 to explore the feasibility of small modular fluoride salt-cooled high temperature reactors (FHRs). A preliminary reactor system concept, SmATHR (for Small modular Advanced High Temperature Reactor) is described, along with an integrated high-temperature thermal energy storage or salt vault system. The SmAHTR is a 125 MWt, integral primary, liquid salt cooled, coated particle-graphite fueled, low-pressure system operating at 700 C. The system employs passive decay heat removal and two-out-of-three , 50% capacity, subsystem redundancy for critical functions. The reactor vessel is sufficiently small to be transportable on standard commercial tractor-trailer transport vehicles. Initial transient analyses indicated the transition from normal reactor operations to passive decay heat removal is accomplished in a manner that preserves robust safety margins at all times during the transient. Numerous trade studies and trade-space considerations are discussed, along with the resultant initial system concept. The current concept is not optimized. Work remains to more completely define the overall system with particular emphasis on refining the final fuel/core configuration, salt vault configuration, and integrated system dynamics and safety behavior.

  16. CLASSIFICATION OF SYSTEMS FOR PASSIVE AFTERHEAT REMOVAL FROM REACTOR CONTAINMENT OF NUCLEAR POWER PLANT WITH WATER-COOLED POWER REACTOR

    Directory of Open Access Journals (Sweden)

    N. Khaled

    2014-01-01

    Full Text Available A classification on systems for passive afterheat removal from reactor containment has been developed in the paper.  The classification permits to make a detailed analysis of various concepts pertaining to systems for passive afterheat removal from reactor containment of new generation. The paper considers main classification features of the given systems.

  17. Advanced gas cooled nuclear reactor materials evaluation and development program. Selection of candidate alloys. Vol. 1. Advanced gas cooled reactor systems definition

    Energy Technology Data Exchange (ETDEWEB)

    Marvin, M.D.

    1978-10-31

    Candidate alloys for a Very High Temperature Reactor (VHTR) Nuclear Process Heal (NPH) and Direct Cycle Helium Turbine (DCHT) applications in terms of the effect of the primary coolant exposure and thermal exposure were evaluated. (FS)

  18. Optimization of power-cycle arrangements for Supercritical Water cooled Reactors (SCWRs)

    Science.gov (United States)

    Lizon-A-Lugrin, Laure

    The world energy demand is continuously rising due to the increase of both the world population and the standard of life quality. Further, to assure both a healthy world economy as well as adequate social standards, in a relatively short term, new energy-conversion technologies are mandatory. Within this framework, a Generation IV International Forum (GIF) was established by the participation of 10 countries to collaborate for developing nuclear power reactors that will replace the present technology by 2030. The main goals of these nuclear-power reactors are: economic competitiveness, sustainability, safety, reliability and resistance to proliferation. As a member of the GIF, Canada has decided to orient its efforts towards the design of a CANDU-type Super Critical Water-cooled Reactor (SCWR). Such a system must run at a coolant outlet temperature of about 625°C and at a pressure of 25 MPa. It is obvious that at such conditions the overall efficiency of this kind of Nuclear Power Plant (NPP) will compete with actual supercritical water-power boilers. In addition, from a heat-transfer viewpoint, the use of a supercritical fluid allows the limitation imposed by Critical Heat Flux (CHF) conditions, which characterize actual technologies, to be removed. Furthermore, it will be also possible to use direct thermodynamic cycles where the supercritical fluid expands right away in a turbine without the necessity of using intermediate steam generators and/or separators. This work presents several thermodynamic cycles that could be appropriate to run SCWR power plants. Improving both thermal efficiency and mechanical power constitutes a multi-objective optimization problem and requires specific tools. To this aim, an efficient and robust evolutionary algorithm, based on genetic algorithm, is used and coupled to an appropriate power plant thermodynamic simulation model. The results provide numerous combinations to achieve a thermal efficiency higher than 50% with a

  19. DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-[TEMPERATURE GAS-COOLED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James; Bayless, Paul; Strydom, Gerhard; Kumar, Akansha; Gougar, Hans

    2016-11-01

    A point design for a graphite-moderated, high-temperature, gas-cooled test reactor (HTG TR) has been developed by Idaho National Laboratory (INL) as part of a United States (U.S.) Department of Energy (DOE) initiative to explore and potentially expand the existing U.S. test reactor capabilities. This paper provides a summary of the design and its main attributes. The 200 MW HTG TR is a thermal-neutron spectrum reactor composed of hexagonal prismatic fuel and graphite reflector blocks. Twelve fuel columns (96 fuel blocks total and 6.34 m active core height) are arranged in two hexagonal rings to form a relatively compact, high-power density, annular core sandwiched between inner, outer, top, and bottom graphite reflectors. The HTG-TR is designed to operate at 7 MPa with a coolant inlet/outlet temperature of 325°C/650°C, and utilizes TRISO particle fuel from the DOE AGR Program with 425 ?m uranium oxycarbide (UCO) kernels and an enrichment of 15.5 wt% 235U. The primary mission of the HTG TR is material irradiation and therefore the core has been specifically designed and optimized to provide the highest possible thermal and fast neutron fluxes. The highest thermal neutron flux (3.90E+14 n/cm2s) occurs in the outer reflector, and the maximum fast flux levels (1.17E+14 n/cm2s) are produced in the central reflector column where most of the graphite has been removed. Due to high core temperatures under accident conditions, all the irradiation test facilities have been located in the inner and outer reflectors where fast flux levels decline. The core features a large number of irradiation positions with large test volumes and long test lengths, ideal for thermal neutron irradiation of large test articles. The total available test volume is more than 1100 liters. Up to four test loop facilities can be accommodated with pressure tube boundaries to isolate test articles and test fluids (e.g., liquid metal, liquid salt, light water) from the helium primary coolant system.

  20. The recent development of fabrication of ODS ferritic steels for supercritical water-cooled reactors core application

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Z.; Li, M.; Liao, L.; Liu, X.; He, P.; Xu, Y.; Chen, W.; Ge, C. [Univ. of Science and Technology Beijing, School of Materials Science and Engineering, Beijing (China)

    2010-07-01

    Development of cladding materials which can work at high temperature is crucial to realize highly efficient and high-burnup operation of Generation IV nuclear energy systems. Oxide dispersion strengthened (ODS) ferritic steel is one of the most promising cladding materials for advanced nuclear reactors, such as supercritical water-cooled reactor. ODS ferritic steels with Cr content of 12, 14 and 18% were designed and fabricated in China through the mechanical alloying (MA) route. The process parameters were discussed and optimized. Mechanical properties were measured at room temperature and high temperature. (author)

  1. Study of thermophysical and thermohydraulic properties of sodium for fast sodium cooled reactors; Estudio de las propiedades termofisicas y termohidraulicas del sodio para reactores rapidos enfriados por sodio

    Energy Technology Data Exchange (ETDEWEB)

    Vega R, A. K.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Gomez T, A. M., E-mail: a.karen.vr@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The importance of liquid sodium lies in its use as a coolant for fast reactors, but why should liquid metal be used as a coolant instead of water? Water is difficult to use as a coolant for a fast nuclear reactor because its acts as a neutron moderator, that is, stop the fast neutrons and converts them to thermal neutrons. Nuclear reactors such as the Pressurized Water Reactor or the Boiling Water Reactor are thermal reactors, which mean they need thermal neutrons for their operation. However, is necessary for fast reactors to conserve as much fast neutrons, so that the liquid metal coolants that do have this capability are implemented. Sodium does not need to be pressurized, its low melting point and its high boiling point, higher than the operating temperature of the reactor, make it an adequate coolant, also has a high thermal conductivity, which is necessary to transfer thermal energy and its viscosity is close to that of the water, which indicates that is an easily transportable liquid and does not corrode the steel parts of the reactor. This paper presents a brief state of the art of the rapid nuclear reactors that operated and currently operate, as well as projects in the door in some countries; types of nuclear reactors which are cooled by liquid sodium and their operation; the mathematical models for obtaining the properties of liquid sodium in a range of 393 to 1673 Kelvin degrees and a pressure atmosphere. Finally a program is presented in FORTRAN named Thermo-Sodium for the calculation of the properties, which requires as input data the Kelvin temperature in which the liquid sodium is found and provides at the user the thermo-physical and thermo-hydraulic properties for that data temperature. Additional to this the user is asked the Reynolds number and the hydraulic diameter in case of knowing them, and in this way the program will provide the value of the convective coefficient and that of the dimensionless numbers: Nusselt, Prandtl and Peclet. (Author)

  2. Safety aspects of fuel behaviour during faults and accidents in pressurised water reactors and in liquid sodium cooled fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.H. (UKAEA Information Services Branch, London); Matthews, J.R. (UKAEA Harwell Lab. (UK). Theoretical Physics Div.); Potter, P.E. (UKAEA Harwell Lab. (UK). Chemistry Div.)

    1989-07-01

    The good safety record of electrical power generating reactors in the European Community is based on a substantial effort to understand the safety characteristics of the reactors and their fuel. In this paper the present state of knowledge of oxide fuels used in current European reactors is reviewed. The main theme of the paper is the importance of the role of fission products and the chemical state of the fuel on all aspects of fuel behaviour. The paper is split into two parts. The first part deals with those aspects specific to water reactors using UO{sub 2} based fuels. The second part of the paper deals with mixed-oxide fuels and the sodium cooled reactors. In each part the following aspects are described: Chemical constitution of the fuel; fuel performance and failure limits; failed fuel behaviour; fuel behaviour in accidents; and the interactions in degraded cores after hypothetical accidents. Future directions of safety related fuel work in Europe are identified. (orig.).

  3. Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook (and others)

    2007-08-15

    This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the

  4. High-temperature gas-cooled reactor safety studies. Progress report for January 1, 1974--June 30, 1975

    Energy Technology Data Exchange (ETDEWEB)

    Cole, T.E.; Sanders, J.P.; Kasten, P.R.

    1977-07-01

    Progress is reported in the following areas: systems and safety analysis; fission product technology; primary coolant technology; seismic and vibration technology; confinement components; primary system materials technology; safety instrumentation; loss of flow accident analysis using HEATUP code; use of coupled-conduction-convection model for core thermal analysis; development of multichannel conduction-convection program HEXEREI; cooling system performance after shutdown; core auxiliary cooling system performance; development of FLODIS code; air ingress into primary systems following DBDA; performance of PCRV thermal barrier cover plates; temperature limits for fuel particle coating failure; tritium distribution and release in HTGR; energy release to PCRV during DBDA; and mathematical models for HTGR reactor safety studies.

  5. Demonstration of leak-before-break in Japan Sodium cooled Fast Reactor (JSFR) pipes

    Energy Technology Data Exchange (ETDEWEB)

    Wakai, Takashi, E-mail: wakai.takashi@jaea.go.jp [Japan Atomic Energy Agency, 4002 Narita-cho, O-arai, Ibaraki 311 1393 (Japan); Machida, Hideo; Yoshida, Shinji [TEPCO Systems Corporation, 2-37-28 Eitai, Koto-ku, Tokyo 135 0034 (Japan); Xu, Yang [Mitsubishi FBR Systems, Inc., 2-34-17 Jingumae, Shibuya-ku, Tokyo 150 0001 (Japan); Tsukimori, Kazuyuki [Japan Atomic Energy Agency, 4002 Narita-cho, O-arai, Ibaraki 311 1393 (Japan)

    2014-04-01

    This paper describes the leak-before-break (LBB) assessment procedure applicable to Japan Sodium cooled Fast Reactor (JSFR) pipes made of modified 9Cr–1Mo steel. For the sodium pipes of JSFR, the continuous leak monitoring will be adopted as an alternative to a volumetric test of the weld joints under conditions that satisfy LBB. Firstly, a LBB assessment flowchart eliminating uncertainty resulted from small scale leakage, such as self plugging phenomenon and influence of crack surface roughness on leak rate, was proposed. Secondly, a rational unstable fracture assessment technique, taking the compliance changing with crack extension into account, was also proposed. Thirdly, a crack opening displacement (COD) assessment technique was developed, because COD assessment method applicable to JSFR pipes – thin wall and small work hardening material – had not been proposed yet. In addition, fracture toughness tests were performed using compact tension (CT) specimens to obtain the fracture toughness, J{sub IC}, and the crack growth resistance (J–R) curve at elevated temperature. Finally, by using the flowchart, proposed techniques and collected data, LBB assessment for the primary sodium pipes of JSFR was conducted. As a result, LBB aspect was successfully demonstrated with sufficient margins.

  6. Analyses of the reflector tank, cold source, and beam tube cooling for ANS reactor

    Energy Technology Data Exchange (ETDEWEB)

    Marland, S. [Tennessee Univ., Knoxville, TN (United States)

    1992-07-01

    This report describes my work as an intern with Martin Marietta Energy Systems, Inc., in the summer of 1991. I was assigned to the Reactor Technology Engineering Department, working on the Advanced Neutron Source (ANS). My first project was to select and analyze sealing systems for the top of the diverter/reflector tank. This involved investigating various metal seals and calculating the forces necessary to maintain an adequate seal. The force calculations led to an analysis of several bolt patterns and lockring concepts that could be used to maintain a seal on the vessel. Another project involved some pressure vessel stress calculations and the calculation of the center of gravity for the cold source assembly. I also completed some sketches of possible cooling channel patterns for the inner vessel of the cold source. In addition, I worked on some thermal design analyses for the reflector tank and beam tubes, including heat transfer calculations and assisting in Patran and Pthermal analyses. To supplement the ANS work, I worked on other projects. I completed some stress/deflection analyses on several different beams. These analyses were done with the aid of CAASE, a beam-analysis software package. An additional project involved bending analysis on a carbon removal system. This study was done to find the deflection of a complex-shaped beam when loaded with a full waste can.

  7. Objective Provision Trees of Reactivity Control Safety Function for Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Bongsuk; Yang, Huichang [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-05-15

    The purpose of this OPT is first to assure the DiD design during the licensing of Sf, but it will also contribute in evaluating the completeness of regulatory requirements under development by Korea Institute of Nuclear Safety (KINS). Based on the definition of Defense-in-Depth (DiD) levels and safety functions for KALIMER Sodium-Cooled Fast Reactor (SFR), suggested in the reference and, Objective Provision Trees (OPTs) of reactivity control function for level 1, 2, 3 and 4 DiD were developed and suggested in this paper. The challenges and mechanisms and provisions were briefly explained in this paper. Comparing the mechanisms and provisions with the requirements will contribute in identifying the missing requirements. Since the design of Prototype Gen-IV Sf (PGSFR) is not mature yet, the OPT is developed for KALIMER design. Developed level 1 to 4 OPTs in this study can be used for the identification of potential design vulnerabilities. When detailed identification of provisions in terms of design features were achieved through the next step of this study, it can contribute to the establishment of defense-in-depth evaluation frame for the regulatory reviews for the licensing process. In the next stage of this study, other safety function will be researched and findings can be suggested as recommendations for the safety improvement.

  8. Corrosion of high temperature alloys in the coolant helium of a gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cabet, C.; Terlain, A. [Service de la Corrosion et du Comportement des Materiaux dans leur Environnement, DEN/DPC - CEA/Saclay, Gif sur Yvette (France); Monnier, A. [Lab. de Genie Electrique de Paris, Plateau du Moulon, Gif sur Yvette (France)

    2004-07-01

    The corrosion of structural alloys in gas cooled reactor environment appears to be a critical issue. The coolant helium proved to contain impurities mainly H{sub 2}, H{sub 2}O, CO, and CH{sub 4} in the microbar range that interact with metallic materials at high temperature. Surface scale formation, bulk carburisation and/or decarburisation can occur, depending on the gas chemistry, the alloy composition and the temperature. These structural transformations can notably influence the component mechanical properties. A short review of the literature on the topic is first given. Corrosion tests with high chromium alloys and a Mo-based alloy were carried out at 750 C in a purposely-designed facility under simulated GCR helium. The first, rather short term, results showed that the Mo-based alloy was inert while the others alloys oxidised during at least 900 hours. The alloy with the higher Al and Ti contents exhibited poor oxidation resistance impeding its use as structural material without further investigations. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, M.L.

    1979-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lindemer, T.B.

    2002-09-05

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

  11. The value of helium-cooled reactor technologies for transmutation of nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, C.; Baxter, A. [General Atomics, Los Alamos, NM (United States)

    2001-07-01

    Helium-cooled reactor technologies offer significant advantages in accomplishing the waste transmutation process. They are ideally suited for use with thermal, epithermal, or fast neutron energy spectra. They can provide a relatively hard thermal neutron spectrum for transmutation of fissionable materials such as Pu-239 using ceramic-coated transmutation fuel particles, a graphite moderator, and a non-fertile burnable poison. These features (1) allow deep levels of transmutation with minimal or no intermediate reprocessing, (2) enhance passive decay heat removal via heat conduction and radiation, (3) allow operation at relatively high temperatures for a highly efficient generation of electricity, and (4) discharge the transmuted waste in a form that is highly resistant to corrosion for long times. They also offer the possibility for the use of epithermal neutrons that can interact with transmutable materials more effectively because of the large atomic cross sections in this energy domain. A fast spectrum may be useful for deep burnup of certain minor actinides. For this application, helium is essentially transparent to neutrons, does not degrade neutron energies, and offers the hardest possible neutron energy environment. In this paper, we report results from recent work on materials transmutation balances, safety, value to a geological repository, and economic considerations. (authors)

  12. Comparative CFD analyses of liquid metal cooled reactor for lunar surface power

    Energy Technology Data Exchange (ETDEWEB)

    Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States)

    2014-12-15

    Highlights: • Performed 3-D thermal-hydraulics analyses of a NaK-56 cooled, sector core reactor for lunar surface power. • Investigated effects mesh refinements and choice of two-layer k–ε and SST k–ω models on results. • Choice of turbulence model slightly affects calculated temperatures and velocities. • Estimates of pressure losses require a finer mesh for convergence than temperature and velocities fields. • SST k–ω model provides more details of mixing eddies, but requires more computation time. - Abstract: This paper presents the results of comparative CFD and thermal-hydraulics analyses of the Solid Core–Sectored Compact Reactor (SC-SCoRe) for lunar surface power. This fast-neutron spectrum, liquid NaK-56 cooled reactor is loaded with highly enriched UN fuel. It nominally generates 1.0 MW{sub th} for ∼21 full-power years at NaK-56 inlet and exit temperatures of 850 K and 900 K. The analyses examine the realizable two-layer k–ε and the Shear Stress Transport (SST) k–ω turbulence models, with different numerical mesh refinements, for simulating the performance of the SC-SCoRe core during nominal operation as well as investigate the effect on the computation time and calculated parameters. In addition, the results calculated for a single tri-lobe flow channel are compared to those obtained using the Detached Eddy Simulation (DES), a hybrid LES and RANS turbulence model. The numerical mesh refinement beyond 2 × 10{sup 7} cells in the flow channels and the choice of turbulence model slightly affect the calculated fuel, core structure, and liquid metal temperatures. They more strongly affect the pressure losses and the intensity of flow mixing and the formation of turbulence eddies in the inlet and exit plenums and exit duct. With the same numerical mesh refinements, the total computation time with the k–ω model is 40–120% longer than with the k-ε model, while the calculated operation parameters are almost identical. The flow

  13. A COOL DUST FACTORY IN THE CRAB NEBULA: A HERSCHEL STUDY OF THE FILAMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Gomez, H. L.; Clark, C. J. R.; Gomez, E. L.; Gear, W. K. [School of Physics and Astronomy, Cardiff University, The Parade, Cardiff CF24 3AA (United Kingdom); Krause, O.; Besel, M.-A.; Bouwman, J.; Henning, Th. [Max-Planck-Institut fuer Astronomie, Koenigstuhl 17, D-69117 Heidelberg (Germany); Barlow, M. J.; Swinyard, B. M.; Owen, P. J.; Matsuura, M. [Department of Physics and Astronomy, University College London, Gower Street, London WC1E 6BT (United Kingdom); Rho, J. [SOFIA Science Center, Universities Space Research Association, NASA Ames Research Center, MS 232, Moffett Field, CA 94035 (United States); Ivison, R. J.; Sibthorpe, B. [UK Astronomy Technology Centre, Royal Observatory Edinburgh, Blackford Hill, Edinburgh EH9 3HJ (United Kingdom); Polehampton, E. T. [Space Science and Technology Department, Rutherford Appleton Laboratory, Oxfordshire, OX11 0QX (United Kingdom)

    2012-11-20

    Whether supernovae are major sources of dust in galaxies is a long-standing debate. We present infrared and submillimeter photometry and spectroscopy from the Herschel Space Observatory of the Crab Nebula between 51 and 670 {mu}m as part of the Mass Loss from Evolved StarS program. We compare the emission detected with Herschel with multiwavelength data including millimeter, radio, mid-infrared, and archive optical images. We carefully remove the synchrotron component using the Herschel and Planck fluxes measured in the same epoch. The contribution from line emission is removed using Herschel spectroscopy combined with Infrared Space Observatory archive data. Several forbidden lines of carbon, oxygen, and nitrogen are detected where multiple velocity components are resolved, deduced to be from the nitrogen-depleted, carbon-rich ejecta. No spectral lines are detected in the SPIRE wavebands; in the PACS bands, the line contribution is 5% and 10% at 70 and 100 {mu}m and negligible at 160 {mu}m. After subtracting the synchrotron and line emission, the remaining far-infrared continuum can be fit with two dust components. Assuming standard interstellar silicates, the mass of the cooler component is 0.24{sup +0.32} {sub -0.08} M {sub Sun} for T = 28.1{sup +5.5} {sub -3.2} K. Amorphous carbon grains require 0.11 {+-} 0.01 M {sub Sun} of dust with T = 33.8{sup +2.3} {sub -1.8} K. A single temperature modified blackbody with 0.14 M {sub Sun} and 0.08 M {sub Sun} for silicate and carbon dust, respectively, provides an adequate fit to the far-infrared region of the spectral energy distribution but is a poor fit at 24-500 {mu}m. The Crab Nebula has condensed most of the relevant refractory elements into dust, suggesting the formation of dust in core-collapse supernova ejecta is efficient.

  14. Feasibility analysis of modified AL-6XN steel for structure component application in supercritical water-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    Xinggang LI; Qingzhi YAN; Rong MA; Haoqiang WANG; Changchun GE

    2009-01-01

    Modified AL-6XN austenite steel was patterned after AL-6XN superaustenitic stainless steel by introducing microalloy elements such as zirconium and titanium in order to adapt to recrystallizing thermo-mechanical treatment and further improve crevice corrosion resistance. Modified AL-6XN exhibited comparable tensile strength, and superior plasticity and impact toughness to commercial AL-6XN steel. The effects of aging behavior on corrosion resistance and impact toughness were measured to evaluate the qualification of modified AL-6XN steel as an in-core component and cladding material in a supercritical water-cooled reactor. Attention should be paid to degradation in corrosion resistance and impact toughness after aging for 50 hours when modified AL-6XN steel is considered as one of the candidate materials for in-core components and cladding tubes in supercritical water-cooled reactors.

  15. Study Neutronic of Small Pb-Bi Cooled Non-Refuelling Nuclear Power Plant Reactor (SPINNOR) with Hexagonal Geometry Calculation

    Science.gov (United States)

    Nur Krisna, Dwita; Su'ud, Zaki

    2017-01-01

    Nuclear reactor technology is growing rapidly, especially in developing Nuclear Power Plant (NPP). The utilization of nuclear energy in power generation systems has been progressing phase of the first generation to the fourth generation. This final project paper discusses the analysis neutronic one-cooled fast reactor type Pb-Bi, which is capable of operating up to 20 years without refueling. This reactor uses Thorium Uranium Nitride as fuel and operating on power range 100-500MWtNPPs. The method of calculation used a computer simulation program utilizing the SRAC. SPINNOR reactor is designed with the geometry of hexagonal shaped terrace that radially divided into three regions, namely the outermost regions with highest percentage of fuel, the middle regions with medium percentage of fuel, and most in the area with the lowest percentage. SPINNOR fast reactor operated for 20 years with variations in the percentage of Uranium-233 by 7%, 7.75%, and 8.5%. The neutronic calculation and analysis show that the design can be optimized in a fast reactor for thermal power output SPINNOR 300MWt with a fuel fraction 60% and variations of Uranium-233 enrichment of 7%-8.5%.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-15

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

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-31

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

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

    Science.gov (United States)

    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

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

    Science.gov (United States)

    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.

  1. CFD Analysis of Turbulent Flow Phenomena in the Lower Plenum of a Prismatic Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    T. Gallaway; S.P. Antal; M.Z. Podowski; D.P. Guillen

    2007-09-01

    This paper is concerned with the implementation of a computational model of turbulent flow in a section of the lower plenum of Very High Temperature Reactor (VHTR). The proposed model has been encoded in a state-of-the-art CFD code, NPHASE. The results of NPHASE predictions have been compared against the experimental data collected using a scaled model of a sub-region in the lower plenum of a modular prismatic gas-cooled reactor. It has been shown that the NPHASE-based model is capable of predicting a three-dimensional velocity field in a complex geometrical configuration of VHTR lower plenum. The current and future validations of computational predictions are necessary for design and analysis of new reactor concepts, as well as for safety analysis and licensing calculations.

  2. THE DIFFERENT EVOLUTION OF GAS AND DUST IN DISKS AROUND SUN-LIKE AND COOL STARS

    NARCIS (Netherlands)

    Pascucci, I.; Apai, D.; Luhman, K.; Henning, Th.; Bouwman, J.; Meyer, M. R.; Lahuis, F.; Natta, A.

    2009-01-01

    Planet formation is profoundly impacted by the properties of protoplanetary disks and their central star. However, how disk properties vary with stellar parameters remains poorly known. Here, we present the first comprehensive, comparative Spitzer/IRS study of the dust and gas properties of disks ar

  3. A 100-kWt NaK-Cooled Space Reactor Concept for an Early-Flight Mission

    Science.gov (United States)

    Poston, David I.

    2003-01-01

    A stainless-steel (SS) sodium-potassium (NaK) cooled reactor could potentially be the first step in utilizing fission technology in space. The sum of all system-level experience for liquid-metal-cooled space reactors has been with NaK, including the SNAP-10a, the only reactor ever launched by the US. This paper describes a 100-kWt NaK reactor, the NaK-100, which is designed to be developed with minimal technical risk. In additional to NaK technology heritage, the NaK-100 uses a proven fuel-form (SS/UO2) and is designed for simplified system integration and testing. The pins are placed within a solid SS prism, and the NaK flows in an annulus between the pins and the prism. The nuclear and thermal-hydraulic performance of the NaK-100 is presented, as well as the major differences between the NaK-100 and SNAP-10a.

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

    Science.gov (United States)

    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.

  5. Application of Simulated Reactivity Feedback in Nonnuclear Testing of a Direct-Drive Gas-Cooled Reactor

    Science.gov (United States)

    Bragg-Sitton, S. M.; Webster, K. L.

    2007-01-01

    Nonnuclear testing can be a valuable tool in the development of an in-space nuclear power or propulsion system. In a nonnuclear test facility, electric heaters are used to simulate heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response and response characteristics, and assess potential design improvements with a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE 100a heat pipe cooled, electrically heated reactor and heat exchanger hardware. This Technical Memorandum discusses the status of the planned dynamic test methodology for implementation in the direct-drive gas-cooled reactor testing and assesses the additional instrumentation needed to implement high-fidelity dynamic testing.

  6. Study on a method for loading a Li compound to produce tritium using high-temperature gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakaya, Hiroyuki, E-mail: nakaya@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Matsuura, Hideaki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Katayama, Kazunari [Department of Advanced Energy Engineering Science, Kyushu University, 6-1 Kasuga-koen, Kasuga 8168580 (Japan); Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki (Japan)

    2015-10-15

    Highlights: • Tritium production by a high-temperature gas-cooled reactor was studied. • The loading method considering tritium outflow suppression was estimated. • A reactor with 600 MWt produced 400–600 g of tritium for 180 days. • A possibility that tritium outflow can be sufficiently suppressed was shown. - Abstract: Tritium production using high-temperature gas-cooled reactors and its outflow from the region loading Li compound into the helium coolant are estimated when considering the suppression of tritium outflow. A Li rod containing a cylindrical Li compound placed in an Al{sub 2}O{sub 3} cladding tube is assumed as a method for loading Li compound. A gas turbine high-temperature reactor of 300 MW electrical nominal capacity (GTHTR300) with 600 MW thermal output power is considered and modeled using the continuous-energy Monte Carlo transport code MVP-BURN, where burn-up simulations are carried out. Tritium outflow is estimated from equilibrium solution for the tritium diffusion equation in the cladding tube. A GTHTR300 can produce 400–600 g of tritium over a 180-day operation using the chosen method of loading the Li compound while minimizing tritium outflow from the cladding tube. Optimizing tritium production while suppressing tritium outflow is discussed.

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

    Science.gov (United States)

    Cisneros, Anselmo Tomas, Jr.

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

  8. Assessment of the Use of Nitrogen Trifluoride for Purifying Coolant and Heat Transfer Salts in the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scheele, Randall D.; Casella, Andrew M.

    2010-09-28

    This report provides an assessment of the use of nitrogen trifluoride for removing oxide and water-caused contaminants in the fluoride salts that will be used as coolants in a molten salt cooled reactor.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  10. Scaling approach and thermal-hydraulic analysis in the reactor cavity cooling system of a high temperature gas -cooled reactor and thermal-jet mixing in a sodium fast reactor

    Science.gov (United States)

    Omotowa, Olumuyiwa A.

    This dissertation develops and demonstrates the application of the top-down and bottom-up scaling methodologies to thermal-hydraulic flows in the reactor cavity cooling system (RCCS) of the high temperature gas reactor (HTGR) and upper plenum of the sodium fast reactor (SFR), respectively. The need to integrate scaled separate effects and integral tests was identified. Experimental studies and computational tools (CFD) have been integrated to guide the engineering design, analysis and assessment of this scaling methods under single and two-phase flow conditions. To test this methods, two applicable case studies are considered, and original contributions are noted. Case 1: "Experimental Study of RCCS for the HTGR". Contributions include validation of scaling analysis using the top-down approach as guide to a ¼-scale integral test facility. System code, RELAP5, was developed based on the derived scaling parameters. Tests performed included system sensitivity to decay heat load and heat sink inventory variations. System behavior under steady-state and transient scenarios were predicted. Results show that the system has the capacity to protect the cavity walls from over-heating during normal operations and provide a means for decay heat removal under accident scenarios. A full width half maximum statistical method was devised to characterize the thermal-hydraulics of the non-linear two-phase oscillatory behavior. This facilitated understanding of the thermal hydraulic coupling of the loop segments of the RCCS, the heat transfer, and the two-phase flashing flow phenomena; thus the impact of scaling overall. Case 2: "Computational Studies of Thermal Jet Mixing in SFR". In the pool-type SFR, susceptible regions to thermal striping are the upper instrumentation structure and the intermediate heat exchanger (IHX). We investigated the thermal mixing above the core to UIS and the potential impact due to poor mixing. The thermal mixing of dual-jet flows at different

  11. Gas-cooled reactor programs: high-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1981

    Energy Technology Data Exchange (ETDEWEB)

    1982-06-01

    Information is presented concerning HTGR chemistry; fueled graphite development; irradiation services for General Atomic Company; prestressed concrete pressure vessel development; HTGR structural materials; graphite development; high-temperature reactor physics studies; shielding studies; component flow test loop studies; core support performance test; and application and project assessments.

  12. Topical report: Natural convection shutdown heat removal test facility (NSTF) evaluation for generating additional reactor cavity cooling system (RCCS) data.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Kilsdonk, D. J.; Tzanos, C.P.; Lomperski, S.; Aeschlimann, R.W.; Pointer, D.; Nuclear Engineering Division

    2005-09-01

    As part of the Department of Energy (DOE) Generation IV roadmapping activity, the Very High Temperature gas cooled Reactor (VHTR) has been selected as the principal concept for hydrogen production and other process-heat applications such as district heating and potable water production. On this basis, the DOE has selected the VHTR for additional R&D with the ultimate goal of demonstrating emission-free electricity and hydrogen production with this advanced reactor concept. One of the key passive safety features of the VHTR is the potential for decay heat removal by natural circulation of air in a Reactor Cavity Cooling System (RCCS). The air-cooled RCCS concept is notably similar to the Reactor Vessel Auxiliary Cooling System (RVACS) that was developed for the General Electric PRISM sodium-cooled fast reactor. As part of the DOE R&D program that supported the development of this fast reactor concept, the Natural Convection Shutdown Heat Removal Test Facility (NSTF) was developed at ANL to provide proof-of-concept data for the RVACS under prototypic natural convection flow, temperature, and heat flux conditions. Due to the similarity between RVACS and the RCCS, current VHTR R&D plans call for the utilization of the NSTF to provide RCCS model development and validation data, in addition to supporting design validation and optimization activities. Both air-cooled and water-cooled RCCS designs are to be included. In support of this effort, ANL has been tasked with the development of an engineering plan for mechanical and instrumentation modifications to NSTF to ensure that sufficiently detailed temperature, heat flux, velocity and turbulence profiles are obtained to adequately qualify the codes under the expected range of air-cooled RCCS flow conditions. Next year, similar work will be carried out for the alternative option of a water-cooled RCCS design. Analysis activities carried out in support of this experiment planning task have shown that: (a) in the RCCS, strong

  13. Validation of Reactor Physics-Thermal hydraulics Calculations for Research Reactors Cooled by the Laminar Flow of Water

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, K. A.; Schubring, D. [Univ. of Florida, Florida (United States); Girardin, G.; Pautz, A. [Swiss Federal Institute of Technology, Zuerich (Switzerland)

    2013-07-01

    A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics

  14. Solid electrolytes for use in lead-bismuth eutectic cooled nuclear reactors

    Science.gov (United States)

    Mariën, A.; Lim, J.; Rosseel, K.; Vandermeulen, W.; Van den Bosch, J.

    2012-08-01

    The operating temperature of electrochemical oxygen sensors can likely be lowered by reducing the solid electrolyte resistance. Most often, these sensors use yttria partially stabilized zirconia (e.g. (Y2O3)0.05(ZrO2)0.95, (5-YSZ)) as solid electrolyte. In this paper, we discuss the use of better conducting ceramics than yttria partially stabilized zirconia, as solid electrolytes for oxygen sensors that can be applied in lead-bismuth eutectic (LBE) cooled nuclear reactors. Two stabilized zirconia ceramics ((Y2O3)0.08(ZrO2)0.92, (8-YSZ); (Sc2O3)0.1(CeO2)0.01(ZrO2)0.89) are investigated as well as a hypostoichiometric perovskite-type La0.8Sr0.2Ga0.8Mg0.2O3-δ. The results of microstructural analyses, thermochemical stability tests in LBE (at 360 °C), as well as mechanical tests and four-probe d.c. conductivity measurements (at 300-800 °C) are discussed and compared with the results that were obtained for a commercially available 5-YSZ (Friatec AG, Germany). Of the three studied ceramics, 8-YSZ was identified as the most promising solid electrolyte to reduce the operating temperature of electrochemical oxygen sensors. http://www.friatec.de/content/friatec/en/Ceramics/FRIALIT-DEGUSSIT-Oxide-Ceramics/downloads/Materials.pdf (this URL was last accessed on February 7th, 2012).

  15. Passive residual energy utilization system in thermal cycles on water-cooled power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Placco, Guilherme M.; Guimaraes, Lamartine N.F., E-mail: placco@ieav.cta.br, E-mail: guimarae@ieav.cta.br [Instituto de Estudos Avancados (IEAV/DCTA) Sao Jose dos Campos, SP (Brazil); Santos, Rubens S. dos, E-mail: rsantos@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN -RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This work presents a concept of a residual energy utilization in nuclear plants thermal cycles. After taking notice of the causes of the Fukushima nuclear plant accident, an idea arose to adapt a passive thermal circuit as part of the ECCS (Emergency Core Cooling System). One of the research topics of IEAv (Institute for Advanced Studies), as part of the heat conversion of a space nuclear power system is a passive multi fluid turbine. One of the main characteristics of this device is its passive capability of staying inert and be brought to power at moments notice. During the first experiments and testing of this passive device, it became clear that any small amount of gas flow would generate power. Given that in the first stages of the Fukushima accident and even during the whole event there was plenty availability of steam flow that would be the proper condition to make the proposed system to work. This system starts in case of failure of the ECCS, including loss of site power, loss of diesel generators and loss of the battery power. This system does not requires electricity to run and will work with bleed steam. It will generate enough power to supply the plant safety system avoiding overheating of the reactor core produced by the decay heat. This passive system uses a modified Tesla type turbine. With the tests conducted until now, it is possible to ensure that the operation of this new turbine in a thermal cycle is very satisfactory and it performs as expected. (author)

  16. Development of Preliminary HT9 Cladding Tube for Sodium-cooled Fast Reactor (SFR)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Baek, Jong Hyuk; Heo, Hyeong Min; Park, Sang Gyu; Kim, Sung Ho; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    To achieve manufacturing technology of the fuel cladding tube in order to keep pace with the predetermined schedule in developing SFR fuel, KAERI has launched in developing fuel cladding tube in cooperation with a domestic steelmaking company. After fabricating medium-sized 1.1 ton HT9 ingot, followed by the multiple processes of hot and cold working, preliminary samples of HT9 seamless cladding tube having 7.4mm in outer diameter, 0.56mm in thickness, and 3m in length were fabricated. The objective of this study is to summarize the brief development status of the HT9 cladding tubes. Mechanical properties like axial tension, biaxial burst, pressurized creep and sodium compatibility of the cladding tubes were carried out to set up the performance evaluation technology to test the prototype FMS cladding tube which is going to be manufactured in next stage. As a part of developing fuel cladding for the Sodium-cooled Fast Reactor (SFR), preliminary HT9 cladding tube was fabricated in cooperation with a domestic steelmaking company. Microstructure as well as mechanical tests like axial tensile test, biaxial burst test, and pressurized creep test of the fuel cladding were carried out. Performance of the domestic HT9 tube was revealed to be similar in the previously fabricated foreign HT9 tube. Further prototype FMS cladding tube is going to be manufactured in next year based on this experience. Various test items like mechanical test, sodium compatibility test, microstructural analysis, basic property, cladding performance under transient situation, and performance under ion and neutron irradiation are going be performed in the future to set up the relevant technology for the licensing of the SFR cladding tube.

  17. Risk management and maintenance optimization of nuclear reactor cooling piping system

    Science.gov (United States)

    Augé, L.; Capra, B.; Lasne, M.; Bernard, O.; Bénéfice, P.; Comby, R.

    2006-11-01

    Seaside nuclear power plants have to face the ageing of nuclear reactor cooling piping systems. In order to minimize the duration of the production unit shutdown, maintenance operations have to be planned well in advance. In a context where owners of infrastructures tend to extend the life span of their goods while having to keep the safety level maximum, Oxand brings its expertise and know-how in management of infrastructures life cycle. A dedicated methodology relies on several modules that all participate in fixing network optimum replacement dates: expertise on ageing mechanisms (corrosion, cement degradation...) and the associated kinetics, expertise on impacts of ageing on functional integrity of piping systems, predictive simulation based on experience feedback, development of monitoring techniques focused on actual threats. More precisely, Oxand has designed a patented monitoring technique based on optic fiber sensors, which aims at controlling the deterioration level of piping systems. This preventive maintenance enables the owner to determine criteria for network replacement based on degradation impacts. This approach helps the owner justify his maintenance strategy and allows him to demonstrate the management of safety level. More generally, all monitoring techniques used by the owners are developed and coupled to predictive simulation tools, notably thanks to processes based on Bayesian approaches. Methodologies to evaluate and optimize operation budgets, depending on predictions of future functional deterioration and available maintenance solutions are also developed and applied. Finally, all information related to infrastructure ageing and available maintenance options are put together to reach the right solution for safe and performing infrastructure management.

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

    Energy Technology Data Exchange (ETDEWEB)

    Gordeev, S.; Hering, W.; Schikorr, M.; Stieglitz, R. [Inst. for Neutron Physic and Reactor Technology, Karlsruhe Inst. of Technology, Campus Nord (Germany)

    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)

  19. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

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

    Science.gov (United States)

    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.

  1. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for FHRs

    Science.gov (United States)

    Lu, Qiuping

    Direct Reactor Auxiliary Cooling System (DRACS) is a passive decay heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines coated particle fuel and a graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three coupled natural circulation/convection loops, relying completely on buoyancy as the driving force. These loops are coupled through two heat exchangers, namely, the DRACS Heat Exchanger (DHX) and the Natural Draft Heat Exchanger (NDHX). In addition, a fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during normal operation of the reactor, but to keep the DRACS ready for activation, if needed, during accidents. While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for FHRs built or tested in the literature. The primary goal of the present research is to design, test, and model the DRACS for FHR applications. Previously, a detailed modular design of the DRACS for a 20-MWth FHR was developed. As a starting point, the DRACS was designed to remove 1% of the reactor nominal power, i.e., 200 kW decay power. In addition, a detailed scaling analysis has been performed to develop the key non-dimensional numbers that characterize the DRACS system. Based on the previous work on the prototypic DRACS design and scaling analysis, two scaled-down test facilities have been designed and constructed, namely, Low-temperature DRACS Test Facility (LTDF) and High-temperature DRACS Test Facility (HTDF). The LTDF has a nominal power capacity of 6 kW. It uses 1.0-MPa water as the primary coolant, 0.1-MPa water as the secondary coolant, and ambient air as the ultimate heat sink. The main purpose of the LTDF is to examine the couplings among the three natural circulation/convection loops in the DRACS, as well as to provide design and operation experience for the HTDF. An extensive test matrix has

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Franck Dechelette

    2014-01-01

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

  4. Analysis of radiological accident emissions of a lead-cooled experimental reactor. LEADER Project; Analisis radiologico de las emisiones en caso de accidente de un reactor experimental refrigerado por plomo. Proyecto LEADER

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Salcedo, F.; Cortes Martin, A.

    2013-07-01

    The LEADER project develops a conceptual level industrial size reactor cooled lead and a demonstration plant of this technology. The project objectives are to define the characteristics and design to installation scale reactor using available technologies and short-term components and assess safety aspects conducting a preliminary analysis of the impact of the facility.

  5. Gas-cooled reactor programs. High-temperature gas-cooled reactor technology development program. Annual progress report, December 31, 1983

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1984-06-01

    ORNL continues to make significant contributions to the national program. In the HTR fuels area, we are providing detailed statistical information on the fission product retention performance of irradiated fuel. Our studies are also providing basic data on the mechanical, physical, and chemical behavior of HTR materials, including metals, ceramics, graphite, and concrete. The ORNL has an important role in the development of improved HTR graphites and in the specification of criteria that need to be met by commercial products. We are also developing improved reactor physics design methods. Our work in component development and testing centers in the Component Flow Test Loop (CFTL), which is being used to evaluate the performance of the HTR core support structure. Other work includes experimental evaluation of the shielding effectiveness of the lower portions of an HTR core. This evaluation is being performed at the ORNL Tower Shielding Facility. Researchers at ORNL are developing welding techniques for attaching steam generator tubing to the tubesheets and are testing ceramic pads on which the core posts rest. They are also performing extensive testing of aggregate materials obtained from potential HTR site areas for possible use in prestressed concrete reactor vessels. During the past year we continued to serve as a peer reviewer of small modular reactor designs being developed by GA and GE with balance-of-plant layouts being developed by Bechtel Group, Inc. We have also evaluated the national need for developing HTRs with emphasis on the longer term applications of the HTRs to fossil conversion processes.

  6. Reactor in search for money. Cooled neutrons for unique research; Reactor zoekt geld. Gekoelde neutronen maken onderzoek mogelijk

    Energy Technology Data Exchange (ETDEWEB)

    Verdult, E.

    2010-12-17

    The modernization of radiation research at the Delft University of Technology depends on subsidies for new instruments. OYSTER (Optimised Yield for Science, Technology and Education of Radiation) is the plan of the Reactor Institute Delft (RID) to realize such a modernization. The article comprises detailed drawings of the inside of the reactor and illustrates the CNIF (Cold Neutron Irradiation Facility) to fight cancer and POSH-PALS (Positron Annihilation Lifetime Spectrometry) to visualize the atomic structure of materials. [Dutch] De modernisering van het stralingsonderzoek aan de Technische Universiteit Delft staat of valt met subsidie voor een nieuwe opzet van het instrumentarium. OYSTER (Optimised Yield for Science, Technology and Education of Radiation) is het plan van het Reactor Instituut Delft (RID) om de installatie te moderniseren. Het artikel bevat gedetailleerde tekeningen van de binnenkant van de reactor en illustreert de CNIF (Cold Neutron Irradiation Facility) voor de bestrijding van kanker en POSH-PALS (Positron Annihilation Lifetime Spectrometry) waarmee de structuur van materialen op atoomniveau inzichtelijk kan worden gemaakt.

  7. Incorporation of statistical distribution of particle properties in chemical reactor design and operation: the cooled tubular reactor

    NARCIS (Netherlands)

    Wijngaarden, R.J.; Westerterp, K.R.

    1992-01-01

    Pellet heat and mass transfer coefficients inside packed beds do not have definite deterministic values, but are stochastic quantities with a certain distribution. Here, a method is presented to incorporate the stochastic distribution of pellet properties in reactor design and operation models. The

  8. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2006-03-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  9. Application of solid dust-free cooling technology in large particles urea cooling%固体无尘冷却技术在大颗粒尿素冷却中的应用

    Institute of Scientific and Technical Information of China (English)

    于泽峰

    2014-01-01

    从冷却原理、设备结构、运行效果等方面详细介绍了固体无尘冷却技术的优点,并与传统的流化床冷却技术进行了经济对比。%The paper detailedly introduces advantages of solid dust-free cooling technology from cooling principle, device structure, and operating results;it also contrasts with traditional fluidized bed cooling technology from economic.

  10. Low-power lead-cooled fast reactor loaded with MOX-fuel

    Science.gov (United States)

    Sitdikov, E. R.; Terekhova, A. M.

    2017-01-01

    Fast reactor for the purpose of implementation of research, education of undergraduate and doctoral students in handling innovative fast reactors and training specialists for atomic research centers and nuclear power plants (BRUTs) was considered. Hard neutron spectrum achieved in the fast reactor with compact core and lead coolant. Possibility of prompt neutron runaway of the reactor is excluded due to the low reactivity margin which is less than the effective fraction of delayed neutrons. The possibility of using MOX fuel in the BRUTs reactor was examined. The effect of Keff growth connected with replacement of natural lead coolant to 208Pb coolant was evaluated. The calculations and reactor core model were performed using the Serpent Monte Carlo code.

  11. Probabilistic Fracture Mechanics Analysis of Boling Water Reactor Vessel for Cool-Down and Low Temperature Over-Pressurization Transients

    Directory of Open Access Journals (Sweden)

    Jeong Soon Park

    2016-04-01

    Full Text Available The failure probabilities of the reactor pressure vessel (RPV for low temperature over-pressurization (LTOP and cool-down transients are calculated in this study. For the cool-down transient, a pressure–temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition (RTNDT. The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

  12. Microbial fouling community analysis of the cooling water system of a nuclear test reactor with emphasis on sulphate reducing bacteria.

    Science.gov (United States)

    Balamurugan, P; Joshi, M Hiren; Rao, T S

    2011-10-01

    Culture and molecular-based techniques were used to characterize bacterial diversity in the cooling water system of a fast breeder test reactor (FBTR). Techniques were selected for special emphasis on sulphate-reducing bacteria (SRB). Water samples from different locations of the FBTR cooling water system, in addition to biofilm scrapings from carbon steel coupons and a control SRB sample were characterized. Whole genome extraction of the water samples and SRB diversity by group specific primers were analysed using nested PCR and denaturing gradient gel electrophoresis (DGGE). The results of the bacterial assay in the cooling water showed that the total culturable bacteria (TCB) ranged from 10(3) to 10(5) cfu ml(-1); iron-reducing bacteria, 10(3) to 10(5) cfu ml(-1); iron oxidizing bacteria, 10(2) to 10(3) cfu ml(-1) and SRB, 2-29 cfu ml(-1). However, the counts of the various bacterial types in the biofilm sample were 2-3 orders of magnitude higher. SRB diversity by the nested PCR-DGGE approach showed the presence of groups 1, 5 and 6 in the FBTR cooling water system; however, groups 2, 3 and 4 were not detected. The study demonstrated that the PCR protocol influenced the results of the diversity analysis. The paper further discusses the microbiota of the cooling water system and its relevance in biofouling.

  13. Design Study of 200MWth Gas Cooled Fast Reactor with Nitride (UN-PuN Fuel Long Life without Refueling

    Directory of Open Access Journals (Sweden)

    Syarifah Ratna Dewi

    2016-01-01

    Full Text Available Design study of 200 MWth Gas Cooled Fast Reactor with UN-PuN fuel long life without refueling has been done. GFR is one type reactor in Generation IV reactor system. It uses helium coolant and fast neutron spectrum. Helium is chemical inert, single phase and low neutron moderation. In this study the calculations are performed by using SRAC code with PIJ calculation for the fuel pin cell calculation and CITATION calculation for core calculation. The data libraries use JENDL 3.2. The variation fuel fractions are 50% until 60%. The diameter active core is 150 cm and the height active core is 100 cm. The reflector radial-axial width is 50 cm. The variation of the powers are 100 MWth up to 500 MWth. The high power causes the high k-eff value. The optimum design is reached when the power is 200 MWth, variation percentage Plutonium for fuel F1:F2:F3=9%:11%:13%. The comparation of fuel:cladding:coolant fraction = 55%:10%:35%. The cooling down time of Plutonium is nine months. The optimum k-eff value is 1.0142 with excess reactivity value 1.403%. The decay of Plutonium decrease k-eff value in the beginning of burn up.

  14. Low-enrichment and long-life Scalable LIquid Metal cooled small Modular (SLIMM-1.2) reactor

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States); Chemical and Biological Engineering Department, University of New Mexico, Albuquerque, NM (United States); Palomino, Luis M.; Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States)

    2017-05-15

    Highlights: • Developed low enrichment and natural circulation cooled SLIMM-1.2 SMR for generating 10–100 MW{sub th}. • Neutronics analyses estimate operation life and temperature reactivity feedback. • At 100 MW{sub th}, SLIMM-1.2 operates for 6.3 FPY without refueling. • SLIMM-1.2 has relatively low power peaking and maximum UN fuel temperature < 1400 K. - Abstract: The Scalable LIquid Metal cooled small Modular (SLIMM-1.0) reactor with uranium nitride fuel enrichment of 17.65% had been developed for generating 10–100 MW{sub th} continuously, without refueling for ∼66 and 5.9 full power years, respectively. Natural circulation of in-vessel liquid sodium (Na) cools the core of this fast energy spectrum reactor during nominal operation and after shutdown, with the aid of a tall chimney and an annular Na/Na heat exchanger (HEX) of concentric helically coiled tubes. The HEX at the top of the downcomer maximizes the static pressure head for natural circulation. In addition to the independent emergency shutdown (RSS) and reactor control (RC), the core negative temperature reactivity feedback safely decreases the reactor thermal power, following modest increases in the temperatures of UN fuel and in-vessel liquid sodium. The decay heat is removed from the core by natural circulation of in-vessel liquid sodium, with aid of the liquid metal heat pipes laid along the reactor vessel wall, and the passive backup cooling system (BCS) using natural circulation of ambient air along the outer surface of the guard vessel wall. This paper investigates modifying the SLIMM-1.0 reactor design to lower the UN fuel enrichment. To arrive at a final reactor design (SLIMM-1.2), the performed neutronics and reactivity depletion analyses examined the effects of various design and material choices on both the cold-clean and the hot-clean excess reactivity, the reactivity shutdown margin, the full power operation life at 100 MW{sub th}, the fissile production and depletion, the

  15. Warm Dust around Cool Stars: Field M Dwarfs with WISE 12 or 22 Micron Excess Emission

    CERN Document Server

    Theissen, Christopher A

    2014-01-01

    Using the SDSS DR7 spectroscopic catalog, we searched the WISE AllWISE catalog to investigate the occurrence of warm dust, as inferred from IR excesses, around field M dwarfs (dMs). We developed SDSS/WISE color selection criteria to identify 175 dMs (from 70,841) that show IR flux greater than typical dM photosphere levels at 12 and/or 22 $\\mu$m, including seven new stars within the Orion OB1 footprint. We characterize the dust populations inferred from each IR excess, and investigate the possibility that these excesses could arise from ultracool binary companions by modeling combined SEDs. Our observed IR fluxes are greater than levels expected from ultracool companions ($>3\\sigma$). We also estimate that the probability the observed IR excesses are due to chance alignments with extragalactic sources is $<$ 0.1%. Using SDSS spectra we measure surface gravity dependent features (K, Na, and CaH 3), and find $<$ 15% of our sample indicate low surface gravities. Examining tracers of youth (H$\\alpha$, UV fl...

  16. Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ferroni, Paolo [Westinghouse Electric Company LLC, Cranberry Township, PA (United States). Global Technology Development; Tatli, Emre [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Czerniak, Luke [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Chien, Hual-Te [Argonne National Lab. (ANL), Argonne, IL (United States); Yoichi, Momozaki [Argonne National Lab. (ANL), Argonne, IL (United States); Bakhtiari, Sasan [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-06-29

    The project “Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems” was conducted jointly by Westinghouse Electric Company (Westinghouse) and Argonne National Laboratory (ANL), over the period October 1, 2013- March 31, 2016. The project’s motivation was the need to provide designers of Sodium Fast Reactors (SFRs) with a validated, state-of-the-art computational tool for the prediction of sodium oxide (Na2O) deposition in small-diameter sodium heat exchanger (HX) channels, such as those in the diffusion bonded HXs proposed for SFRs coupled with a supercritical CO2 (sCO2) Brayton cycle power conversion system. In SFRs, Na2O deposition can potentially occur following accidental air ingress in the intermediate heat transport system (IHTS) sodium and simultaneous failure of the IHTS sodium cold trap. In this scenario, oxygen can travel through the IHTS loop and reach the coldest regions, represented by the cold end of the sodium channels of the HXs, where Na2O precipitation may initiate and continue. In addition to deteriorating HX heat transfer and pressure drop performance, Na2O deposition can lead to channel plugging especially when the size of the sodium channels is small, which is the case for diffusion bonded HXs whose sodium channel hydraulic diameter is generally below 5 mm. Sodium oxide melts at a high temperature well above the sodium melting temperature such that removal of a solid plug such as through dissolution by pure sodium could take a lengthy time. The Sodium Plugging Phenomena Loop (SPPL) was developed at ANL, prior to this project, for investigating Na2O deposition phenomena within sodium channels that are prototypical of the diffusion bonded HX channels envisioned for SFR-sCO2 systems. In this project, a Computational Fluid Dynamic (CFD) model capable of simulating the thermal-hydraulics of the SPPL test

  17. Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco, E-mail: gianfranco.caruso@uniroma1.it

    2016-08-15

    Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

  18. Scaled Facility Design Approach for Pool-Type Lead-Bismuth Eutectic Cooled Small Modular Reactor Utilizing Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sangrok; Shin, Yong-Hoon; Lee, Jueun; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    2015-10-15

    In low carbon era, nuclear energy is the most prominent energy source of electricity. For steady ecofriendly nuclear energy supply, Generation IV reactors which are future nuclear reactor require safety, sustainability, economics and non-proliferation as four criteria. Lead cooled fast reactor (LFR) is one of these reactor type and Generation IV international forum (GIF) adapted three reference LFR systems which are a small and movable systems with long life without refueling, intermediate size and huge electricity generation system for power grid. NUTRECK (Nuclear Transmutation Energy Center of Korea) has been designed reactor called URANUS (Ubiquitous, Rugged, Accident-forgiving, Non-proliferating, and Ultra-lasting Sustainer) which is small modular reactor and using lead-bismuth eutectic coolant. To prove natural circulation capability of URANUS and analyze design based accidents, scaling mock-up experiment facility will be constructed. In this paper, simple specifications of URANUS will be presented. Then based on this feature, scaling law and scaled facility design results are presented. To validate safety feature and thermodynamics characteristic of URANUS, scaled mockup facility of URANUS is designed based on the scaling law. This mockup adapts two area scale factors, core and lower parts of mock-up are scaled for 3D flow experiment. Upper parts are scaled different size to reduce electricity power and LBE tonnage. This hybrid scaling method could distort some thermal-hydraulic parameters, however, key parameters for experiment will be matched for up-scaling. Detailed design of mock-up will be determined through iteration for design optimization.

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

    Science.gov (United States)

    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.

  20. Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rohanda, Anis [Center for Reactor Technology and Nuclear Safety – BATAN Kawasan PUSPIPTEK Gd. No. 80 Serpong, Tangerang Selatan 15310 (Indonesia); Waris, Abdul [Physics Department of ITB, Indonesia anis-rohanda@yahoo.com (Indonesia)

    2015-04-16

    Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on {sup 16}O(n,p){sup 16}N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

  1. THE INVESTIGATION OF BURNUP CHARACTERISTICS USING THE SERPENT MONTE CARLO CODE FOR A SODIUM COOLED FAST REACTOR

    Directory of Open Access Journals (Sweden)

    MEHMET E. KORKMAZ

    2014-06-01

    Full Text Available In this research, we investigated the burnup characteristics and the conversion of fertile 232Th into fissile 233U in the core of a Sodium-Cooled Fast Reactor (SFR. The SFR fuel assemblies were designed for burning 232Th fuel (fuel pin 1 and 233U fuel (fuel pin 2 and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method and TTA (Transmutation Trajectory Analysis method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff was between 0.964 and 0.954 and peaking factor is 1.88867.

  2. The investigation of burnup characteristics using the serpent Monte Carlo code for a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, Mehmet E.; Agar, Osman [Karamanoglu Mehmetbey University, Faculty of Kamil Oezdag Science, Karaman (Turkmenistan)

    2014-06-15

    In this research, we investigated the burnup characteristics and the conversion of fertile {sup 232}Th into fissile {sup 233}U in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning {sup 232}Th fuel (fuel pin 1) and {sup 233}U fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.

  3. In vessel detection of delayed neutron emitters from clad failure in sodium cooled nuclear reactors: An estimation of the signal

    Science.gov (United States)

    Filliatre, P.; Jammes, C.; Chapoutier, N.; Jeannot, J.-P.; Jadot, F.; Batail, R.; Verrier, D.

    2014-04-01

    The detection of clad failures is mandatory in sodium-cooled fast neutron reactors in compliance with the "clean sodium" concept. An in-vessel detection system, sensitive to delayed neutrons from fission products released into the primary coolant by failures, partially tested in SUPERPHENIX, is foreseen in current SFR projects in order to reduce significantly the delay before an alarm is issued. In this paper, an estimation of the signal received by such a system in case of a failure is derived, taking the French project ASTRID as a working example. This failure induced signal is compared to that of the contribution of the neutrons from the core itself. The sensitivity of the system is defined in terms of minimal detectable surface of clad failure. Possible solutions to improve this sensitivity are discussed, involving either the sensor itself, or the hydraulic design of the vessel in the early stage of the reactor conception.

  4. The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts

    Science.gov (United States)

    Scarlat, Raluca O.; Peterson, Per F.

    2014-01-01

    The fluoride salt cooled high temperature reactor (FHR) is a class of fission reactor designs that use liquid fluoride salt coolant, TRISO coated particle fuel, and graphite moderator. Heavy ion fusion (HIF) can likewise make use of liquid fluoride salts, to create thick or thin liquid layers to protect structures in the target chamber from ablation by target X-rays and damage from fusion neutron irradiation. This presentation summarizes ongoing work in support of design development and safety analysis of FHR systems. Development work for fluoride salt systems with application to both FHR and HIF includes thermal-hydraulic modeling and experimentation, salt chemistry control, tritium management, salt corrosion of metallic alloys, and development of major components (e.g., pumps, heat exchangers) and gas-Brayton cycle power conversion systems. In support of FHR development, a thermal-hydraulic experimental test bay for separate effects (SETs) and integral effect tests (IETs) was built at UC Berkeley, and a second IET facility is under design. The experiments investigate heat transfer and fluid dynamics and they make use of oils as simulant fluids at reduced scale, temperature, and power of the prototypical salt-cooled system. With direct application to HIF, vortex tube flow was investigated in scaled experiments with mineral oil. Liquid jets response to impulse loading was likewise studied using water as a simulant fluid. A set of four workshops engaging industry and national laboratory experts were completed in 2012, with the goal of developing a technology pathway to the design and licensing of a commercial FHR. The pathway will include experimental and modeling efforts at universities and national laboratories, requirements for a component test facility for reliability testing of fluoride salt equipment at prototypical conditions, requirements for an FHR test reactor, and development of a pre-conceptual design for a commercial reactor.

  5. Neutron flux measurements in the side-core region of Hunterston B advanced gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Allen, D.A. [Serco, Rutherford House, Quedgeley, Gloucester, GL2 4NF (United Kingdom); Shaw, S.E. [British Energy, Barnett Way, Barnwood, Gloucester, GL4 3RS (United Kingdom); Huggon, A.P.; Steadman, R.J.; Thornton, D.A. [Serco, Rutherford House, Quedgeley, Gloucester, GL2 4NF (United Kingdom); Whiley, G.S. [British Energy, Barnett Way, Barnwood, Gloucester, GL4 3RS (United Kingdom)

    2011-07-01

    The core restraints of advanced gas-cooled reactors are important structural components that are required to maintain the geometric integrity of the cores. A review of neutron dosimetry for the sister stations Hunterston B and Hinkley Point B identified that earlier conservative assessments predicted high thermal neutron dose rates to key components of the restraint structure (the restraint rod welds), with the implication that some of them may be predicted to fail during a seismic event. A revised assessment was therefore undertaken [Thornton, D. A., Allen, D. A., Tyrrell, R. J., Meese, T. C., Huggon, A.P., Whiley, G. S., and Mossop, J. R., 'A Dosimetry Assessment for the Core Restraint of an Advanced Gas Cooled Reactor,' Proceedings of the 13. International Symposium on Reactor Dosimetry (ISRD-13, May 2008), World Scientific, River Edge, NJ, 2009, W. Voorbraak, L. Debarberis, and P. D'hondt, Eds., pp. 679-687] using a detailed 3D model and a Monte Carlo radiation transport program, MCBEND. This reassessment resulted in more realistic fast and thermal neutron dose recommendations, the latter in particular being much lower than had been thought previously. It is now desirable to improve confidence in these predictions by providing direct validation of the MCBEND model through the use of neutron flux measurements. This paper describes the programme of work being undertaken to deploy two neutron flux measurement 'stringers' within the side-core region of one of the Hunterston B reactors for the purpose of validating the MCBEND model. The design of the stringers and the determination of the preferred deployment locations have been informed by the use of detailed MCBEND flux calculations. These computational studies represent a rare opportunity to design a flux measurement beforehand, with the clear intention of minimising the anticipated uncertainties and obtaining measurements that are known to be representative of the neutron fields to which

  6. Solar energy as an alternate energy source to mixed oxide fuels in light-water cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bertini, H.W.

    1977-06-30

    Supplemental information pertaining to the generic environmental impact statement on the Pu recycling process for mixed oxide light-water cooled reactors (GESMO) was requested from several sources. In particular, the role of alternate sources of energy was to be explored and the implications of these alternate sources to the question of Pu recycle in LWRs were to be investigated. In this vein, solar energy as an alternate source is the main subject of this report, along with other information related to solar energy. The general conclusion is that solar energy should have little effect on the decisions concerning GESMO.

  7. Development of dual cooled annular fuel and its possibility to enhance both economy and safety of light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yanghyun; Kim, Keonsik; Park, Jeongyong; Yang, Yongsik; Kim, Hyungkyu; In, Wangkee; Song, Kunwoo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-03-15

    Over the past few decades, extensive studies have been performed to improve the reliability and safety of light water reactor (LWR) fuel. In recent years, power updating of about 10% is being achieved by modifying safety analysis methodology and subsequent increase in safety margin. But departure from nucleate boiling (DNB) and loss of coolant accident (LOCA) are still two of the most important limiting factors which would restrict power updating more than 10%. Duel cooled annular fuel, cooled in both internal and external cooling channel, has advantages of considerably lower heat flux and lower fuel temperature than conventional solid fuel. While lower heat flus gives higher DNB margin for the same power retie, lower temperature reduces the stored energy of fuel. However, there are many technical issues that should be addressed before any new type of fuel can be considered for application to LWR. This paper describes the key technologies that Korea Atomic Energy Research Institute (KAERI) has developed for dual cooled annular fuel and discusses the feasibility of its application to LWR.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Science.gov (United States)

    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.

  10. Review of selected aspects of the Army Gas-Cooled Reactor Systems Program

    Energy Technology Data Exchange (ETDEWEB)

    None

    1965-08-27

    Information is presented concerning the AGCRS program; ML-1 reactor skid refurbishing program; ML-1-IM fabrication status; power conversion system component testing program; ML-1 demonstration test program; and applications of ML-1 technology.

  11. ICP-MS analysis of fission product diffusion in graphite for High-Temperature Gas-Cooled Reactors

    Science.gov (United States)

    Carter, Lukas M.

    Release of radioactive fission products from nuclear fuel during normal reactor operation or in accident scenarios is a fundamental safety concern. Of paramount importance are the understanding and elucidation of mechanisms of chemical interaction, nuclear interaction, and transport phenomena involving fission products. Worldwide efforts to reduce fossil fuel dependence coupled with an increasing overall energy demand have generated renewed enthusiasm toward nuclear power technologies, and as such, these mechanisms continue to be the subjects of vigorous research. High-Temperature Gas-Cooled Reactors (HTGRs or VHTRs) remain one of the most promising candidates for the next generation of nuclear power reactors. An extant knowledge gap specific to HTGR technology derives from an incomplete understanding of fission product transport in major core materials under HTGR operational conditions. Our specific interest in the current work is diffusion in reactor graphite. Development of methods for analysis of diffusion of multiple fission products is key to providing accurate models for fission product release from HTGR core components and the reactor as a whole. In the present work, a specialized diffusion cell has been developed and constructed to facilitate real-time diffusion measurements via ICP-MS. The cell utilizes a helium gas-jet system which transports diffusing fission products to the mass spectrometer using carbon nanoparticles. The setup was designed to replicate conditions present in a functioning HTGR, and can be configured for real-time release or permeation measurements of single or multiple fission products from graphite or other core materials. In the present work, we have analyzed release rates of cesium in graphite grades IG-110, NBG-18, and a commercial grade of graphite, as well as release of iodine in IG-110. Additionally we have investigated infusion of graphite samples with Cs, I, Sr, Ag, and other surrogate fission products for use in release or

  12. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

    Energy Technology Data Exchange (ETDEWEB)

    Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

    2005-02-13

    The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

  13. Modeling of natural circulation for the inherent safety analysis of sodium cooled fast reactors

    Directory of Open Access Journals (Sweden)

    A.S. Bochkarev

    2016-12-01

    Full Text Available The paper discusses a set of developed integrated one-dimensional models of thermal-hydraulic processes that contribute to the removal of decay heat in a BN-type reactor. The assumptions and constraints involved in one-dimensional equations of unsteady natural convection in closed circuits have been analyzed. It has been shown that the calculated values of the primary circuit sodium temperature and flow rate in conditions with a loss of heat sink and with a forced circulation of the primary coolant are in a reasonable agreement with the results of a benchmark experiment in the PHENIX reactor. The model makes it possible to assess the effects general thermophysical and geometrical parameters and the selected technology have on the efficiency of passive heat removal by the natural coolant convection in the reactor tank and in the emergency heat removal system's intermediate circuit and by the heat transfer through the reactor vessel. The model is a part of an integrated algorithm used to assess the inherent safety level of advanced fast neutron reactors and is intended primarily to develop, at the early conceptual design stage, the recommendations and requirements with respect to the reactor equipment parameters leading to an increase in the reactor inherent safety. The model will be used to identify the set of quantitative thermal-hydraulic criteria that have an effect on the dynamics of emergency transients leading to a potential loss of integrity by the reactor safety barriers, and to formulate such limits for the defined criteria as would cause, if observed, the requirement for the safety barrier integrity to be met under any combination of the accident initiating events.

  14. Simulation of In-Vessel Corium Retention through External Reactor Vessel Cooling for SMART using SIMPLE

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jin-Sung; Son, Donggun; Park, Rae-Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Thermal load analysis from the corium pool to the outer reactor vessel in the lower plenum of the reactor vessel is necessary to evaluate the effect of the IVR-ERVC during a severe accident for SMART. A computational code called SIMPLE (Sever Invessel Melt Progression in Lower plenum Environment) has been developed for analyze transient behavior of molten corium in the lower plenum, interaction between corium and coolant, and heat-up and ablation of reactor vessel wall. In this study, heat load analysis of the reactor vessel for SMART has been conducted using the SIMPLE. Transient behavior of the molten corium in the lower plenum and IVR-ERVC for SMART has been simulated using SIMPLE. Heat flux from the corium pool to the outer reactor vessel is concentrated in metallic layer by the focusing effect. As a result, metallic layer shows higher temperature than the oxidic layer. Also, vessel wall of metallic layer has been ablated by the high in-vessel temperature. Ex-vessel temperature of the metallic layer was maintained 390 K and vessel thickness was maintained 14 cm. It means that the reactor vessel integrity is maintained by the IVR-ERVC.

  15. In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

    Energy Technology Data Exchange (ETDEWEB)

    F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

    2005-01-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

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

    Science.gov (United States)

    Mella, R.; Wenman, M. R.

    2013-06-01

    Thermo-mechanical contributions to pellet-clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS's well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used. The usability of a FE based fuel performance code would be an enhancement over past codes. Pre- and post-processors have lowered the entry barrier for the development of a fuel performance model to permit the ability to model complicated systems. Typical runtimes for a 5 year axisymmetric model takes less than one hour on a single core workstation. The current model has implemented: Non-linear fuel thermal behaviour, including a complex description of heat flow in the fuel. Coupled with a variety of

  17. INVESTIGATION OF FUNDAMENTAL THERMAL-HYDRAULIC PHENOMENA IN ADVANCED GAS-COOLED REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    INVESTIGATION OF FUNDAMENTAL THERMAL-HYDRAULIC PHE

    2006-09-01

    INL LDRD funded research was conducted at MIT to experimentally characterize mixed convection heat transfer in gas-cooled fast reactor (GFR) core channels in collaboration with INL personnel. The GFR for Generation IV has generated considerable interest and is under development in the U.S., France, and Japan. One of the key candidates is a block-core configuration first proposed by MIT, has the potential to operate in Deteriorated Turbulent Heat Transfer (DTHT) regime or in the transition between the DTHT and normal forced or laminar convection regime during post-loss-of-coolant accident (LOCA) conditions. This is contrary to most industrial applications where operation is in a well-defined and well-known turbulent forced convection regime. As a result, important new need emerged to develop heat transfer correlations that make possible rigorous and accurate predictions of Decay Heat Removal (DHR) during post LOCA in these regimes. Extensive literature review on these regimes was performed and a number of the available correlations was collected in: (1) forced laminar, (2) forced turbulent, (3) mixed convection laminar, (4) buoyancy driven DTHT and (5) acceleration driven DTHT regimes. Preliminary analysis on the GFR DHR system was performed and using the literature review results and GFR conditions. It confirmed that the GFR block type core has a potential to operate in the DTHT regime. Further, a newly proposed approach proved that gas, liquid and super critical fluids all behave differently in single channel under DTHT regime conditions, thus making it questionable to extrapolate liquid or supercritical fluid data to gas flow heat transfer. Experimental data were collected with three different gases (nitrogen, helium and carbon dioxide) in various heat transfer regimes. Each gas unveiled different physical phenomena. All data basically covered the forced turbulent heat transfer regime, nitrogen data covered the acceleration driven DTHT and buoyancy driven DTHT

  18. Development of Core Heat Removal Objective Provision Trees for Sodium-Cooled Fast Reactor Defense-in-Depth Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Huichang; Kang, Bongsuk; Lee, Youngho [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    Based on the definition of Defense-in-Depth levels and safety functions for KALIMER sodium-cooled fast reactor, suggested in the reference and, OPTs for level 1, 2, and 3 defense-in-depth and core heat removal safety function, were developed and suggested in this paper. The purpose of this OPT is first to assure the defensein-depth design during the licensing of Sodium-Cooled Fast Reactors (SFR), but it will also contribute in evaluating the completeness of regulatory requirements under development by Korea Institute of Nuclear Safety (KINS). The challenges and mechanisms and provisions were briefly explained in this paper. Comparing the mechanisms and provisions with the requirements will contribute in identifying the missing requirements. Since the design of PGSFR (Prototype Gen-IV SFR) is not mature yet, the OPT is developed for KALIMER design. Developed OPTs in this study can be used for the identification of potential design vulnerabilities. When detailed identification of provisions in terms of design features were achieved through the next step of this study, it can contribute to the establishment of defensein-depth evaluation frame for the regulatory reviews for the licensing process. At this moment, the identified provisions have both aspects as requirements and design features already adopted in KALIMER design. In the next stage of this study, derived provisions to be adopted will be compared with the actual design features and findings can be suggested as recommendations for the safety improvement.

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

    Science.gov (United States)

    Rostamian, Maziar

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

  20. Sensitivity Analysis of Fuel Centerline Temperatures in SuperCritical Water-cooled Reactors (SCWRs)

    Science.gov (United States)

    Abdalla, Ayman

    SuperCritical Water-cooled Reactors (SCWRs) are one of the six nuclear-reactor concepts currently being developed under the Generation-IV International Forum (GIF). A main advantage of SCW Nuclear Power Plants (NPPs) is that they offer higher thermal efficiencies compared to those of current conventional NPPs. Unlike today's conventional NPPs, which have thermal efficiencies between 30 - 35%, SCW NPPs will have thermal efficiencies within a range of 45 - 50%, owing to high operating temperatures and pressures (i.e., coolant temperatures as high as 625°C at 25 MPa pressure). The use of current fuel bundles with UO2 fuel at the high operating parameters of SCWRs may cause high fuel centerline temperatures, which could lead to fuel failure and fission gas release. Studies have shown that when the Variant-20 (43-element) fuel bundle was examined at SCW conditions, the fuel centerline temperature industry limit of 1850°C for UO2 and the sheath temperature design limit of 850°C might be exceeded. Therefore, new fuel-bundle designs, which comply with the design requirements, are required for future use in SCWRs. The main objective of this study to conduct a sensitivity analysis in order to identify the main factors that leads to fuel centerline temperature reduction. Therefore, a 54-element fuel bundle with smaller diameter of fuel elements compared to that of the 43-element bundle was designed and various nuclear fuels are examined for future use in a generic Pressure Tube (PT) SCWR. The 54-element bundle consists of 53 heated fuel elements with an outer diameter of 9.5 mm and one central unheated element of 20-mm outer diameter which contains burnable poison. The 54-element fuel bundle has an outer diameter of 103.45 mm, which is the same as the outer diameter of the 43-element fuel bundle. After developing the 54-element fuel bundle, one-dimensional heat-transfer analysis was conducted using MATLAB and NIST REFPROP programs. As a result, the Heat Transfer

  1. Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ferroni, Paolo [Westinghouse Electric Company LLC, Cranberry Township, PA (United States). Global Technology Development; Tatli, Emre [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Czerniak, Luke [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Chien, Hual-Te [Argonne National Lab. (ANL), Argonne, IL (United States); Yoichi, Momozaki [Argonne National Lab. (ANL), Argonne, IL (United States); Bakhtiari, Sasan [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-06-29

    The project “Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems” was conducted jointly by Westinghouse Electric Company (Westinghouse) and Argonne National Laboratory (ANL), over the period October 1, 2013- March 31, 2016. The project’s motivation was the need to provide designers of Sodium Fast Reactors (SFRs) with a validated, state-of-the-art computational tool for the prediction of sodium oxide (Na2O) deposition in small-diameter sodium heat exchanger (HX) channels, such as those in the diffusion bonded HXs proposed for SFRs coupled with a supercritical CO2 (sCO2) Brayton cycle power conversion system. In SFRs, Na2O deposition can potentially occur following accidental air ingress in the intermediate heat transport system (IHTS) sodium and simultaneous failure of the IHTS sodium cold trap. In this scenario, oxygen can travel through the IHTS loop and reach the coldest regions, represented by the cold end of the sodium channels of the HXs, where Na2O precipitation may initiate and continue. In addition to deteriorating HX heat transfer and pressure drop performance, Na2O deposition can lead to channel plugging especially when the size of the sodium channels is small, which is the case for diffusion bonded HXs whose sodium channel hydraulic diameter is generally below 5 mm. Sodium oxide melts at a high temperature well above the sodium melting temperature such that removal of a solid plug such as through dissolution by pure sodium could take a lengthy time. The Sodium Plugging Phenomena Loop (SPPL) was developed at ANL, prior to this project, for investigating Na2O deposition phenomena within sodium channels that are prototypical of the diffusion bonded HX channels envisioned for SFR-sCO2 systems. In this project, a Computational Fluid Dynamic (CFD) model capable of simulating the thermal-hydraulics of the SPPL test

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

    Science.gov (United States)

    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.

  3. An Assessment of Fission Product Scrubbing in Sodium Pools Following a Core Damage Event in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, M.; Farmer, M.; Grabaskas, D.

    2017-06-26

    The U.S. Nuclear Regulatory Commission has stated that mechanistic source term (MST) calculations are expected to be required as part of the advanced reactor licensing process. A recent study by Argonne National Laboratory has concluded that fission product scrubbing in sodium pools is an important aspect of an MST calculation for a sodium-cooled fast reactor (SFR). To model the phenomena associated with sodium pool scrubbing, a computational tool, developed as part of the Integral Fast Reactor (IFR) program, was utilized in an MST trial calculation. This tool was developed by applying classical theories of aerosol scrubbing to the decontamination of gases produced as a result of postulated fuel pin failures during an SFR accident scenario. The model currently considers aerosol capture by Brownian diffusion, inertial deposition, and gravitational sedimentation. The effects of sodium vapour condensation on aerosol scrubbing are also treated. This paper provides details of the individual scrubbing mechanisms utilized in the IFR code as well as results from a trial mechanistic source term assessment led by Argonne National Laboratory in 2016.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors. A review. Pt. I. Key areas

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, Nilay K. [Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamilnadu (India). Dept. of Atomic Energy (DAE); Raj, Baldev [P.S. Govindaswamy Naidu (PSG) Institutions, Coimbatore, Tamilnadu (India)

    2013-11-15

    Identification of development areas and their implementation for rotatable plug (RP) inflatable seals of Na cooled, 500 Mw (e) Prototype Fast Breeder Reactor (PFBR) and 40 MW (t) Fast Breeder Test Reactor (FBTR) are described, largely based on a late 1990s survey of cover gas seal development (1950s - early 1990s) which defined a set of shortlisted design options and developmental strategy to minimize effort, cost and time. Comparative studies of top shield sealing and evolving FBR designs suggest suitability of inflatable seal as primary barrier in RPs. International experience identified choice and qualification of seal elastomer under synergistic degrading environment of reactor as the prime element of development. The low pressure, non-reinforced, unbeaded, PFBR inflatable seal (made of 50/50 blend of Viton {sup registered} GBL 200S/600S) developed for 10 y life provides a unification scheme for nuclear elastomeric sealing based on 5 peroxide cured fluoroelastomer blend formulations, 1 finite element analysis approach, 1 Teflon-like plasma coating technique and 2 manufacturing processes promising significant gains in standardization, economy and safety. Uniqueness was ab initio development in the absence of established industry or ready-made supply. Part I addresses key areas of design shortlisting, strategy, development and unification with a backdrop of international evolution. (orig.)

  6. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, H. Oh, PhD; Cliff Davis; Richard Moore

    2004-11-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 900 degrees C or operational fuel temperatures above 1250 degrees C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR's higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Engineering and Environmental Laboratory (INEEL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world's computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertaninty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Spitzer SAGE-Spec: Near Infrared Spectroscopy, Dust Shells, and Cool Envelopes in Extreme Large Magellanic Cloud AGB Stars

    CERN Document Server

    Blum, R D; Kemper, F; Ling, B; Volk, K

    2014-01-01

    K-band spectra are presented for a sample of 39 Spitzer IRS SAGE-Spec sources in the Large Magellanic Cloud. The spectra exhibit characteristics in very good agreement with their positions in the near infrared - Spitzer color-magnitude diagrams and their properties as deduced from the Spitzer IRS spectra. Specifically, the near infrared spectra show strong atomic and molecular features representative of oxygen-rich and carbon-rich asymptotic giant branch stars, respectively. A small subset of stars were chosen from the luminous and red extreme "tip" of the color magnitude diagram. These objects have properties consistent with dusty envelopes but also cool, carbon-rich "stellar" cores. Modest amounts of dust mass loss combine with the stellar spectral energy distribution to make these objects appear extreme in their near infrared and mid infrared colors. One object in our sample, HV 915, a known post asymptotic giant branch star of the RV Tau type exhibits CO 2.3 micron band head emission consistent with previ...

  10. Gas-Cooled Thermal Reactor Program. Semiannual technical progress report, April 1, 1983-September 30, 1983

    Energy Technology Data Exchange (ETDEWEB)

    1983-12-01

    An assessment of the HTGR opportunities from the year 2000 through 2045 was the principal activity on the Market Definition Task (WBS 03). Within the Plant Technology (WBS 13) task, there were activities to develop analytical methods for investigation of Coolant Transport Behavior and to define methods and criteria for High Temperature Structural Engineering design. The activities in support of the HTGR-SC/C Lead Plant (WBS 30 and 31) were the participation in the Lead Plant System Engineering (LPSE) effort and the plant simulation task. The efforts on the Advanced HTGR systems was performed under the Modular Reactor Systems (MRS) (WBS 41) to study the potential for multiple small reactors to provide lower costs, improved safety, and higher availability than the large monolithic core reactors.

  11. Production of liquid fuels with a high-temperature gas-cooled reactor

    Science.gov (United States)

    Quade, R. N.; Vrable, D. L.; Green, L., Jr.

    An exploration is made of the technical, economic and environmental impact feasibility of integrating coal liquefaction methods directly and indirectly with a nuclear reactor source of process heat, with stress on the production of synthetic jet fuel. Production figures and operating costs are compared for indirect conventional and nuclear processes using Lurgi-Fischer-Tropsch technology with direct conventional and nuclear techniques employing the advanced SRC-II technology, and it is concluded that significant advantages in coal savings and environmental impact can be expected from nuclear reactor integration.

  12. Pilot-plant testing of magnetic filters for the N-Reactor primary cooling circuit

    Energy Technology Data Exchange (ETDEWEB)

    Emory, B.B.

    1982-01-01

    Data obtained during the laboratory loop test program using the high power HGMF indicates that removal efficiency for /sup 60/Co and subsequently the bulk of the crud, will be greater than 90% at field strength above .1 Tesla for the expanded metal mesh matrix. However, since /sup 54/Mn seems to exhibit paramagnetic behavior and the possibility of quantities of alpha iron forming during reactor shut down from oxygen inleakage, a field strength of .5 to 1 Tesla may be more appropriate for a full scale on-reactor installation. Crud loading of 50 gm per kg of matrix weight are readily obtainable and up to twice that amount has been reached.

  13. Study of hydrogen generation plant coupled to high temperature gas cooled reactor

    Science.gov (United States)

    Brown, Nicholas Robert

    Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. While both high temperature nuclear reactors and hydrogen generation plants have high individual degrees of development, study of the coupled plant is lacking. Particularly absent are considerations of the transient behavior of the coupled plant, as well as studies of the safety of the overall plant. The aim of this document is to contribute knowledge to the effort of nuclear hydrogen generation. In particular, this study regards identification of safety issues in the coupled plant and the transient modeling of some leading candidates for implementation in the Nuclear Hydrogen Initiative (NHI). The Sulfur Iodine (SI) and Hybrid Sulfur (HyS) cycles are considered as candidate hydrogen generation schemes. Several thermodynamically derived chemical reaction chamber models are coupled to a well-known reference design of a high temperature nuclear reactor. These chemical reaction chamber models have several dimensions of validation, including detailed steady state flowsheets, integrated loop test data, and bench scale chemical kinetics. Eight unique case studies are performed based on a thorough literature review of possible events. The case studies are: (1) feed flow failure from one section of the chemical plant to another, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without SCRAM, (6) total failure of the chemical plant, (7) parametric study of the temperature in an individual reaction chamber, and (8) control rod insertion in the nuclear reactor. Various parametric

  14. The European JASMIN Project for the Development of a New Safety Simulation Code, ASTEC-Na, for Na-cooled Fast Neutron Reactors

    OpenAIRE

    GIRAULT N.; VAN DORSSELAERE J.p.; Jacq, F.; BRILLANT G.; KISSANE Martin; BANDINI, G; Buck,M.; CHAMPIGNY J.; Hering, W; Perez-Martin, S.; Herranz, L; RAISON Philippe; Reinke, N; TUCEK Kamil; VERWAERDE D.

    2012-01-01

    The 4-year JASMIN collaborative project, involving 9 organizations, was launched by IRSN end of 2011 within the 7th European R&D Framework Programme on the enhancement of Na-cooled Fast Neutron Reactors (SFR) safety for a higher resistance to severe accidents. The project aims at developing a new European simulation code, ASTEC-Na, with a modern architecture, sufficiently flexible to account for innovative reactor designs and eventually new types of fuel and claddings and accounting for resul...

  15. A simplified Probabilistic Safety Assesment of a Steam-Methane Reforming Hydrogen Production Plant coupled to a High-Temperature Gas Cooled Nuclear Reactor

    OpenAIRE

    Nelson Edelstein, Pamela; Flores Flores, Alain; Francois Lacouture, Juan Luis

    2005-01-01

    A Probabilistic Safety Assessment (PSA) is being developed for a steam-methane reforming hydrogen production plant linked to a High-Temperature Gas Cooled Nuclear Reactor (HTGR). This work is based on the Japan Atomic Energy Research Institute’s (JAERI) High Temperature Test Reactor (HTTR) prototype in Japan. This study has two major objectives: calculate the risk to onsite and offsite individuals, and calculate the frequency of different types of damage to the complex. A simplified HAZOP...

  16. Safety properties of sodium-cooled fast reactors%钠冷快堆及其安全特性

    Institute of Scientific and Technical Information of China (English)

    徐銤; 杨红义

    2016-01-01

    钠冷快堆是第四代核能系统国际论坛(GIF)公布的6种第四代先进反应堆中研发进展最快、最接近满足商业核电厂需要的堆型。钠冷快堆因其在固有安全性以及可增殖核燃料、嬗变长寿命放射性废物等方面的优势,得到了世界各国的重视。文章以中国第一座钠冷快堆——中国实验快堆(China Experimental Fast Reactor,CEFR)为例,介绍了钠冷快堆在设计及运行方面的安全特性。%The sodium-cooled fast reactor is the fastest prototype and the closest to com-mercialization for nuclear power plants amongst the six types of fourth generation reactors, as an-nounced at the Generation IV International Forum. Many countries are paying more and more at-tention to the research and development of these reactors, due to the inherent safety features, effi-cient utilization of uranium with the breeding of the plutonium, and transmutation of long-lived ac-tinides. The design and operational safety characteristics of the China Experimental Fast Reactor are reviewed in this paper.

  17. Saturated Adaptive Output-Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2014-11-01

    Full Text Available Small modular reactors (SMRs are those nuclear fission reactors with electrical output powers of less than 300 MWe. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear plants with high safety-level and economical competitive power. Power-level control is crucial in providing grid-appropriation for all types of SMRs. Usually, there exists nonlinearity, parameter uncertainty and control input saturation in the SMR-based plant dynamics. Motivated by this, a novel saturated adaptive output-feedback power-level control of the MHTGR is proposed in this paper. This newly-built control law has the virtues of having relatively neat form, of being strong adaptive to parameter uncertainty and of being able to compensate control input saturation, which are given by constructing Lyapunov functions based upon the shifted-ectropies of neutron kinetics and reactor thermal-hydraulics, giving an online tuning algorithm for the controller parameters and proposing a control input saturation compensator respectively. It is proved theoretically that input-to-state stability (ISS can be guaranteed for the corresponding closed-loop system. In order to verify the theoretical results, this new control strategy is then applied to the large-range power maneuvering control for the MHTGR of the HTR-PM plant. Numerical simulation results show not only the relationship between regulating performance and control input saturation bound but also the feasibility of applying this saturated adaptive control law practically.

  18. A preliminary approach to the extension of the Transuranus code to the fuel rod performance analysis of HLM-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Luzzi, L.; Botazzoli, P.; Devita, M.; Di Marcello, V.; Pastore, G. [Department of Energy, Politecnico di Milano, Enrico Fermi Center for Nuclear Studies - CeSNEF, via Ponzio 34/3, 20133 Milano (Italy)

    2010-07-01

    This paper briefly presents a preliminary modelling approach, aimed at the extension of the TRANSURANUS code to the fuel rod performance analysis of Heavy Liquid Metal (HLM) cooled nuclear reactors, with specific reference to the employment of the T91 steel as cladding material and of the liquid Lead-Bismuth Eutectic (LBE) as coolant. On the basis of literature indications, correlations for heat transfer to LBE, corrosion behaviour and thermo-mechanical properties of T91 are proposed, and some open issues are discussed in prospect of more reliable fuel rod performance analysis of HLM-cooled nuclear reactors. (authors)

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

    Science.gov (United States)

    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.

  20. Optimization of Overall Project of ChinaLEad-Alloy Cooled Reactor

    Institute of Scientific and Technical Information of China (English)

    DUAN; Tian-ying; LONG; Bin; HU; Yun; LIU; Xing-ming; SUN; Gang; LIU; Yi-zhe; YANG; Yong; PEI; Zhi-yong; FENG; Wei-wei; YU; Xiao

    2012-01-01

    <正>The application of accelerator driven sub-critical reactor which is able to transmute spent fuel and long-lived radioactive fission products is highly concerned in the international research projects. Lots of researches have been done in many countries. While China Institute of Atomic Energy has built a fast

  1. High Temperature Fusion Reactor Cooling Using Brayton Cycle Based Partial Energy Conversion

    Science.gov (United States)

    Juhasz, Albert J.; Sawicki, Jerzy T.

    2003-01-01

    For some future space power systems using high temperature nuclear heat sources most of the output energy will be used in other than electrical form, and only a fraction of the total thermal energy generated will need to be converted to electrical work. The paper describes the conceptual design of such a partial energy conversion system, consisting of a high temperature fusion reactor operating in series with a high temperature radiator and in parallel with dual closed cycle gas turbine (CCGT) power systems, also referred to as closed Brayton cycle (CBC) systems, which are supplied with a fraction of the reactor thermal energy for conversion to electric power. Most of the fusion reactor's output is in the form of charged plasma which is expanded through a magnetic nozzle of the interplanetary propulsion system. Reactor heat energy is ducted to the high temperature series radiator utilizing the electric power generated to drive a helium gas circulation fan. In addition to discussing the thermodynamic aspects of the system design the authors include a brief overview of the gas turbine and fan rotor-dynamics and proposed bearing support technology along with performance characteristics of the three phase AC electric power generator and fan drive motor.

  2. Dynamic simulation of a sodium-cooled fast reactor power plant

    Energy Technology Data Exchange (ETDEWEB)

    Shinaishin, M.A.M.

    1976-08-01

    Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is the subject of this dissertation. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. A simulator referred to as the CRBRP-DCHT simulator was constructed for studying transients due to unexpected accidents followed by reactor scram. In this simulator emphasis was placed on simulating the auxiliary heat removal system, in order to determine its capability to remove the after-shut down fission and decay heat. The transients studied using the two versions of the ATWS simulator include step and ramp reactivity perturbations, and an electrical load perturbation in the controlled plant. An uncontrolled control rod withdrawal followed by reactor scram was studied using the DCHT simulator, although the duration of this transient was restricted to 20 sec. because of computer limitations. The results agree very well with the expected physical behavior of the plant.

  3. Literature review on metallic fuel source term for sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Nam Duk; Bae, Moo Hoon; Shin, An Dong; Huh, Chang Wook [Korea Institute of Nuclear Safety, Daejon (Korea, Republic of)

    2012-10-15

    Source term is defined as the release of radionuclides from the fuel and coolant into the containment, and subsequently to the environment, following a severe accident where a significant portion of the reactor core has melted. 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. Apart from assessing the radiological consequences for siting, it is also important for designing filtering systems and even reactor components. Overly conservative source term for light water reactor, TID 14844 demands for very fast closure of main steam isolation valves, rapid startup of emergency diesels, and safety systems designed to mitigate gaseous iodine. In spite of this importance, most of the knowledge we have for SFR source term comes from the research performed before 1980s. Moreover, majority of the work on metallic fuels was done during the late 1950's through the 1960's. This paper reviews and summarizes the main characteristics of SFR source terms based on the available literatures.

  4. Sustainability and Efficiency Improvements of Gas-Cooled High Temperature Reactors

    NARCIS (Netherlands)

    Marmier, A.

    2012-01-01

    The work presented in this thesis covers three fundamental aspects of High Temperature Reactor (HTR) performance, namely fuel testing under irradiation for maximized safety and sustainability, fuel architecture for improved economy and sustainability, and a novel Balance of Plant concept to enable f

  5. Porosity Effect in the Core Thermal Hydraulics for Ultra High Temperature Gas-cooled Reactor

    Directory of Open Access Journals (Sweden)

    Motoo Fumizawa

    2008-12-01

    Full Text Available This study presents an experimental method of porosity evaluation and a predictive thermal-hydraulic analysis with packed spheres in a nuclear reactor core. The porosity experiments were carried out in both a fully shaken state with the closest possible packing and in a state of non-vibration. The predictive analysis considering the fixed porosity value was applied as a design condition for an Ultra High Temperature Reactor Experiment (UHTREX. The thermal-hydraulic computer code was developed and identified as PEBTEMP. The highest outlet coolant temperature of 1316 oC was achieved in the case of an UHTREX at Los Alamos Scientific Laboratory, which was a small scale UHTR. In the present study, the fuel was changed to a pebble type, a porous media. In order to compare the present pebble bed reactor and UHTREX, a calculation based on HTGR-GT300 was carried out in similar conditions with UHTREX; in other words, with an inlet coolant temperature of 871oC, system pressure of 3.45 MPa and power density of 1.3 w/cm3. As a result, the fuel temperature in the present pebble bed reactor showed an extremely lower value compared to that of UHTREX.

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

    Science.gov (United States)

    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.

  7. Sustainability and Efficiency Improvements of Gas-Cooled High Temperature Reactors

    NARCIS (Netherlands)

    Marmier, A.

    2012-01-01

    The work presented in this thesis covers three fundamental aspects of High Temperature Reactor (HTR) performance, namely fuel testing under irradiation for maximized safety and sustainability, fuel architecture for improved economy and sustainability, and a novel Balance of Plant concept to enable

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-15

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

  9. CFD simulations of moderator flow inside Calandria of the Passive Moderator Cooling System of an advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pal, Eshita [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Joshi, Jyeshtharaj B., E-mail: jbjoshi@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Department of Chemical Engineering, Institute of Chemical Technology, Matunga, Mumbai 400019 India (India); Nayak, Arun K. [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Vijayan, Pallippattu K., E-mail: vijayanp@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India)

    2015-10-15

    Highlights: • CFD simulations in the Calandria of an advanced reactor under natural circulation. • Under natural convection, majority of the flow recirculates within the Calandria. • Maximum temperature is located at the top and center of the fuel channel matrix. • During SBO, temperature inside Calandria is stratified. - Abstract: Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario (Kumar et al., 2013). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria

  10. Review of core disruptive accident analysis for liquid-metal cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. C.; Na, B. C.; Hahn, D. H

    1997-04-01

    Analysis methodologies of core disruptive accidents (CDAs) are reviewed. The role of CDAS in the overall safety evaluation of fast reactors has not always been well defined nor universally agreed upon. However, they have become a traditional issue in LMR safety, design, and licensing. The study is for the understanding of fast reactor behavior under CDA conditions to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features for the KALIMER developments. The methods used to analyze CDAs from initiating event to complete core disruption are described. Two examples of CDA analyses for CRBRP and ALMR are given and R and D needed for better understanding of CDA phenomena are proposed. (author). 10 refs., 2 tabs., 3 figs

  11. Design of a Low Power, Fast-Spectrum, Liquid-Metal Cooled Surface Reactor System

    Science.gov (United States)

    Marcille, T. F.; Dixon, D. D.; Fischer, G. A.; Doherty, S. P.; Poston, D. I.; Kapernick, R. J.

    2006-01-01

    In the current 2005 US budget environment, competition for fiscal resources make funding for comprehensive space reactor development programs difficult to justify and accommodate. Simultaneously, the need to develop these systems to provide planetary and deep space-enabling power systems is increasing. Given that environment, designs intended to satisfy reasonable near-term surface missions, using affordable technology-ready materials and processes warrant serious consideration. An initial lunar application design incorporating a stainless structure, 880 K pumped NaK coolant system and a stainless/UO2 fuel system can be designed, fabricated and tested for a fraction of the cost of recent high-profile reactor programs (JIMO, SP-100). Along with the cost reductions associated with the use of qualified materials and processes, this design offers a low-risk, high-reliability implementation associated with mission specific low temperature, low burnup, five year operating lifetime requirements.

  12. Annular core for Modular High-Temperature Gas-Cooled Reactor (MHTGR)

    Energy Technology Data Exchange (ETDEWEB)

    Turner, R.F.; Baxter, A.M.; Stansfield, O.M.; Vollman, R.E.

    1987-08-01

    The active core of the 350 MW(t) MHTGR is annular in configuration, shaped to provide a large external surface-to-volume ratio for the transport of heat radially to the reactor vessel in case of a loss of coolant flow. For a given fuel temperature limit, the annular core provides approximately 40% greater power output over a typical cylindrical configuration. The reactor core is made up of columns of hexagonal blocks, each 793-mm high and 360-mm wide. The active core is 3.5 m in o.d., 1.65 m in i.d., and 7.93-m tall. Fuel elements contain TRISO-coated microspheres of 19.8% enriched uranium oxycarbide and of fertile thorium oxide. The core is controlled by 30 control rods which enter the inner and outer side reflectors from above.

  13. Annular core for the Modular High-Temperature Gas-cooled Reactor (MHTGR)

    Energy Technology Data Exchange (ETDEWEB)

    Turner, R.F.; Baxter, A.M.; Stansfield, O.M.; Vollman, R.E.

    The active core of the 350 MW(t) MHTGR is annular in configuration, shaped to provide a large external surface-to-volume ratio for the transport of heat radially to the reactor vessel in case of a loss of coolant flow. For a given fuel temperature limit, the annular core provides approximately 40% greater power output over a typical cylindrical configuration. The reactor core is made up to columns of hexagonal blocks, each 793 mm high and 360 mm wide. The active core is 3.5 m in outside diameter, 1.65 m in inside diameter, and 7.93 m tall. Fuel elements contain TRISO-coated microspheres of 19.8% enriched uranium oxycarbide and of fertile thorium oxide. The core is controlled by 30 control rods which enter the inner and outer side reflectors from above.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-09-15

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

  15. Analysis on thermophoretic deposit of fine particle on water wall of 10 MW high temperature gas-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    ZHOU Tao; YANG Rui-Chang; JIA Dou-Nan

    2005-01-01

    The water wall is an important part of the passive natural circulation residual heat removal system in a high temperature gas-cooled reactor. The maximum temperatures of the pressure shell and the water wall are calculated using annular vertical closed cavity model. Fine particles can deposit on the water wall due to the thermophore sis effect. This deposit can affect heat transfer. The thermophoretic deposit efficiency is calculated by using Batch and Shen's formula fitted for both laminar flow and turbulent flow. The calculated results indicate that natural convection is turbulent in the closed cavity. The transient thermophoretic deposit efficiency rises with the increase of the pressure shell's temperature. Its maximum value is 14%.

  16. Modelling 3D crack propagation in ageing graphite bricks of Advanced Gas-cooled Reactor power plant

    Directory of Open Access Journals (Sweden)

    Thi-Tuyet-Giang Vo

    2015-10-01

    Full Text Available In this paper, crack propagation in Advanced Gas-cooled Reactor (AGR graphite bricks with ageing properties is studied using the eXtended Finite Element Method (X-FEM. A parametric study for crack propagation, including the influence of different initial crack shapes and propagation criteria, is conducted. The results obtained in the benchmark study show that the crack paths from X-FEM are similar to the experimental ones. The accuracy of the strain energy release rate computation in a heterogeneous material is also evaluated using a finite difference approach. Planar and non-planar 3D crack growth simulations are presented to demonstrate the robustness and the versatility of the method utilized. Finally, this work contributes to the better understanding of crack propagation behaviour in AGR graphite bricks and so contributes to the extension of the AGR plants’ lifetimes in the UK by reducing uncertainties.

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

    Science.gov (United States)

    Kuk, Seoung Woo; Kim, Ki Hwan; Kim, Jong Hwan; Song, Hoon; Oh, Seok Jin; Park, Jeong-Yong; Lee, Chan Bock; Youn, Young-Sang; Kim, Jong-Yun

    2017-04-01

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

  18. Study of the Effect of Reduced Iron Temperature Rising on Total Carbon Formation in Iron Reactor Isobaric and Cooling Zone

    Directory of Open Access Journals (Sweden)

    Bayu Alamsari

    2010-01-01

    Full Text Available We presented the mathematical model in the iron reactor. The model was limited to Isobaric Zone and Cooling Zone termed as IZ and CZ, respectively. The simulation was done by adapting the heat and mass transfer equations. The main purpose of this paper is to estimate the temperature increasing effect of reduced-iron on sponge-iron quality. The calculations are solved using Finite Element Method (FEM. The results showed that the temperature and concentration values from the simulation have high similarity to the reference data with Root-Mean-Square Error (RMSE about 0.7. The formation of total-carbon in the both zones decreased metallization degree until 1.72%. The increase in reduced-iron temperature higher than 1200 K produces total-carbon higher than 3%. Thus the increase in reduced-iron temperature more than 1200 K is not recommended because it can decrease metallization degree.

  19. Studies on disintegrating spherical fuel elements of high temperature gas-cooled reactor by a electrochemical method

    Science.gov (United States)

    Tian, Lifang; Wen, Mingfen; Chen, Jing

    2013-01-01

    Spherical fuel elements of a high temperature gas-cooled reactor were disintegrated through a electrochemical method with NaNO3 as electrolyte. The X-ray diffraction spectra and total carbon contents of the graphite fragments were determined, and the results agreed with those from simulated fuel elements. After conducting the characterization analysis and the leaching experiment of coated fuel particles, the uranium concentrations of leaching solutions and spent electrolyte were found to be at background levels. The results demonstrate the effectiveness of the improved electrochemical method with NaNO3 as electrolyte in disintegrating the unirradiated fuel elements without any damage to the coated fuel particles. Moreover, the method avoided unexpected radioactivity contamination to the graphite matrix and spent electrolyte.

  20. Analysis of two-phase flow instability in helical tube steam generator in high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Yu; Lv, Xuefeng; Wang, Shengfei; Niu, Fenglei; Tian, Li [North China Electric Power Univ., Beijing (Switzerland)

    2012-03-15

    The steam generator composed of multi-helical tubes is used in high temperature gas cooled reactors and two-phase flow instability should be avoided in design. And density-wave oscillation which is mainly due to flow, density and the relationship between the pressure drop delays and feedback effects is one of the two-phase flow instability phenomena easily to occur. Here drift-flux model is used to simulate the performance of the fluid in the secondary side and frequency domain and time domain methods are used to evaluate whether the density-wave oscillation will happen or not. Several operating conditions with nominal power from 15% to 30% are calculated in this paper. The results of the two methods are in accordance, flow instability will occur when power is less than 20% nominal power, which is also according with the result of the experiments well.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-09-15

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

  2. Analysis of dashpot performance for rotating control drums of a lithium cooled fast reactor concept

    Science.gov (United States)

    Wenzler, C. J.

    1972-01-01

    A dashpot was incorporated in the design of the drive train of the rotating control drum to prevent shock damage to the control drum and drive train at the termination of a scram action. A rotating vane dashpot using reactor coolant lithium as a damping fluid appears to be the best candidate of the various damping devices explored. A performance analysis, results and discussion of vane type dashpots are presented.

  3. Discovery of "Warm Dust" Galaxies in Clusters at z~0.3: Evidence for Stripping of Cool Dust in the Dense Environment?

    CERN Document Server

    Rawle, T D; Egami, E; Chung, S M; Pérez-González, P G; Smail, I; Walth, G; Altieri, B; Appleton, P; Alba, A Berciano; Blain, A W; Dessauges-Zavadsky, M; Fadda, D; Gonzalez, A H; Pereira, M J; Valtchanov, I; van der Werf, P P; Zemcov, M

    2012-01-01

    Using far-infrared imaging from the "Herschel Lensing Survey", we derive dust properties of spectroscopically-confirmed cluster member galaxies within two massive systems at z~0.3: the merging Bullet Cluster and the more relaxed MS2137.3-2353. Most star-forming cluster sources (~90%) have characteristic dust temperatures similar to local field galaxies of comparable infrared (IR) luminosity (T_dust ~ 30K). Several sub-LIRG (L_IR 37K) with far-infrared spectral energy distribution (SED) shapes resembling LIRG-type local templates. X-ray and mid-infrared data suggest that obscured active galactic nuclei do not contribute significantly to the infrared flux of these "warm dust" galaxies. Sources of comparable IR-luminosity and dust temperature are not observed in the relaxed cluster MS2137, although the significance is too low to speculate on an origin involving recent cluster merging. "Warm dust" galaxies are, however, statistically rarer in field samples (> 3sigma), indicating that the responsible mechanism ma...

  4. Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors. Publishable Final Activity Report

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C., E-mail: kuijper@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Somers, J.; Van Den Durpel, L.; Chauvet, V.; Cerullo, N.; Cetnar, J.; Abram, T.; Bakker, K.; Bomboni, E.; Bernnat, W.; Domanska, J.G.; Girardi, E.; De Haas, J.B.M.; Hesketh, K.; Hiernaut, J.P.; Hossain, K.; Jonnet, J.; Kim, Y.; Kloosterman, J.L.; Kopec, M.; Murgatroyd, J.; Millington, D.; Lecarpentier, D.; Lomonaco, G.; McEachern, D.; Meier, A.; Mignanelli, M.; Nabielek, H.; Oppe, J.; Petrov, B.Y.; Pohl, C.; Ruetten, H.J.; Schihab, S.; Toury, G.; Trakas, C.; Venneri, F.; Verfondern, K.; Werner, H.; Wiss, T.; Zakova, J.

    2010-11-15

    The PUMA project -the acronym stands for 'Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors'- was a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Program (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO{sub 2}-free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR

  5. Homogenization of the internal structures of a reactor with the cooling fluid

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F. [CEA Saclay, SEMT, 91 - Gif sur Yvette (France); Bliard, F. [Socotec Industrie, Service AME, 78 - Montigny le Bretonneux (France)

    2001-07-01

    To take into account the influence of a structure net among a fluid flow, without modelling exactly the structure shape, a concept of ''equivalent porosity method'' was developed. The structures are considered as solid pores inside the fluid. The structure presence is represented by three parameters: a porosity, a shape coefficient and a pressure loss coefficient. The method was studied for an Hypothetical Core Disruptive Accident in a Liquid Metal Fast Breeder Reactor, but it can be applied to any problem involving fluid flow getting through a solid net. The model was implemented in the computer code CASTEM-PLEXUS and validated on an analytical shock tube test, simulating an horizontal slice of a schematic LMFBR in case of a HCDA (bubble at high pressure, liquid sodium and internal structures of the reactor). A short parametric study shows the influence of the porosity and the structure shape on the pressure wave impacting the shock tube bottom. These results were used to simulate numerically the HCDA mechanical effects in a small scale reactor mock-up. (author)

  6. Implementation Plan for Qualification of Sodium-Cooled Fast Reactor Technology Information

    Energy Technology Data Exchange (ETDEWEB)

    Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Honma, George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    This document identifies and discusses implementation elements that can be used to facilitate consistent and systematic evaluation processes relating to quality attributes of technical information (with focus on SFR technology) that will be used to support licensing of advanced reactor designs. Information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The approach for determining acceptability of test data, analysis, and/or other technical information is based on guidance provided in INL/EXT-15-35805, “Guidance on Evaluating Historic Technology Information for Use in Advanced Reactor Licensing.” The implementation plan can be adopted into a working procedure at each of the national laboratories performing data qualification, or by applicants seeking future license application for advanced reactor technology.

  7. Bio-processing of copper from combined smelter dust and flotation concentrate: A comparative study on the stirred tank and airlift reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vakylabad, Ali Behrad, E-mail: alibehzad86@yahoo.co.uk [Department of Mining Engineering, Shahid Bahonar University, Kerman (Iran, Islamic Republic of); Engineers of Nano and Bio Advanced Sciences Company (ENBASCo.), ATIC, Mohaghegh University (Iran, Islamic Republic of); Schaffie, Mahin [Department of Chemical Engineering, Shahid Bahonar University of Kerman, Kerman (Iran, Islamic Republic of); Mineral Industries Research Centre (MIRC), Shahid Bahonar University of Kerman, Kerman (Iran, Islamic Republic of); Ranjbar, Mohammad [Department of Mining Engineering, Shahid Bahonar University, Kerman (Iran, Islamic Republic of); Mineral Industries Research Centre (MIRC), Shahid Bahonar University of Kerman, Kerman (Iran, Islamic Republic of); Manafi, Zahra [Sarcheshmeh Copper Complex, National Iranian Copper Industry Company (Iran, Islamic Republic of); Darezereshki, Esmaeel [Mineral Industries Research Centre (MIRC), Shahid Bahonar University of Kerman, Kerman (Iran, Islamic Republic of); Energy and Environmental Engineering Research Center (EERC), Shahid Bahonar University of Kerman, Kerman (Iran, Islamic Republic of)

    2012-11-30

    Highlights: Black-Right-Pointing-Pointer Flotation concentrate and smelter dust were sampled and combined. Black-Right-Pointing-Pointer Copper bioleaching from the combined was investigated. Black-Right-Pointing-Pointer Two bio-reactors were investigated and optimized: stirred and airlift. Black-Right-Pointing-Pointer STRs had better technical conditions and situations for bacterial leaching. - Abstract: To scrutinize the influence of the design and type of the bioreactors on the bioleaching efficiency, the bioleaching were evaluated in a batch airlift and a batch stirred tank bioreactors with mixed mesophilic and mixed moderately thermophilic bacteria. According to the results, maximum copper recoveries were achieved using the cultures in the stirred tank bioreactors. It is worth noting that the main phase of the flotation concentrate was chalcopyrite (as a primary sulphide), but the smelter dust mainly contained secondary copper sulphides such as Cu{sub 2}S, CuS, and Cu{sub 5}FeS{sub 4}.Under optimum conditions, copper dissolution from the combined flotation concentrate and smelter dust (as an environmental hazard) reached 94.50% in the STR, and 88.02% in the airlift reactor with moderately thermophilic, after 23 days. Also, copper extractions calculated for the bioleaching using mesophilic bacteria were 48.73% and 37.19% in the STR (stirred tank reactor) and the airlift bioreactor, respectively. In addition, the SEM/EDS, XRD, chemical, and mineralogical analyses and studies confirmed the above results.

  8. On0Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ayman I. Hawari; Mohamed A. Bourham

    2010-04-22

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

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

    Directory of Open Access Journals (Sweden)

    Sidi-Ali Kamel

    2013-01-01

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

  10. Stability analysis of the Korean prototype generation-IV sodium-cooled fast reactor using linear frequency domain approach

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Ji; Ha, Pham Nhu Viet; Lim, Jae Yong; Hahn, Do Hee; Kang, Chang Mu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of). Fast Reactor Development Div.

    2016-03-15

    The Korea Atomic Energy Research Institute (KAERI) has been developing the 150 MWe Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR). The design concept is highly based on passive safety mechanisms, minimizing the need for engineered safety systems. Presently, it is of primary importance to assure the reactor dynamics and stability against small reactivity disturbances under power operating conditions. KAERI has therefore developed the NuSTAB code for stability analysis of the PGSFR. In NuSTAB, the neutron-kinetic and thermal-hydraulic coupling equations are linearized to form the characteristic equation, which is solved as a generalized eigenvalue problem for determining the decay ratio, an indicator of the system stability. In this paper, the stability of the PGSFR was analyzed by applying the point kinetic and spatial kinetic options in the NuSTAB code. System responses to temperature feedbacks including the Doppler effect, thermal expansion, coolant density change, and overall feedback were studied. The results indicate that the initial U and final TRU cores of the PGSFR are both inherently stable thanks to the temperature feedbacks.

  11. An Artificial Neural Network Compensated Output Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2014-02-01

    Full Text Available Small modular reactors (SMRs could be beneficial in providing electricity power safely and also be viable for applications such as seawater desalination and heat production. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear power plants. Since the MHTGR dynamics display high nonlinearity and parameter uncertainty, it is necessary to develop a nonlinear adaptive power-level control law which is not only beneficial to the safe, stable, efficient and autonomous operation of the MHTGR, but also easy to implement practically. In this paper, based on the concept of shifted-ectropy and the physically-based control design approach, it is proved theoretically that the simple proportional-differential (PD output-feedback power-level control can provide asymptotic closed-loop stability. Then, based on the strong approximation capability of the multi-layer perceptron (MLP artificial neural network (ANN, a compensator is established to suppress the negative influence caused by system parameter uncertainty. It is also proved that the MLP-compensated PD power-level control law constituted by an experientially-tuned PD regulator and this MLP-based compensator can guarantee bounded closed-loop stability. Numerical simulation results not only verify the theoretical results, but also illustrate the high performance of this MLP-compensated PD power-level controller in suppressing the oscillation of process variables caused by system parameter uncertainty.

  12. Experimental Study of the Effect of Graphite Dispersion on the Heat Transfer Phenomena in a Reactor Cavity Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Vaghetto, Rodolfo; Capone, Luigi; Hassan, Yassin A

    2011-05-31

    An experimental activity was performed to observe and study the effects of graphite dispersion and deposition on thermal-hydraulic phenomena in a reactor cavity cooling system (RCCS). The small-scale RCCS experimental facility (16.5 x 16.5 x 30.4 cm) used for this activity represents half of the reactor cavity with an electrically heated vessel. Water flowing through five vertical pipes removes the heat produced in the vessel and releases it into the environment by mixing with cold water in a large tank. The particle image velocimetry technique was used to study the velocity field of the air inside the cavity. A set of 52 thermocouples was installed in the facility to monitor the temperature profiles of the vessel, pipe walls, and air. Ten grams of a fine graphite powder (average particle size 2 m) was injected into the cavity through a spraying nozzle placed at the bottom of the vessel. The temperatures and air velocity field were recorded and compared with the measurements obtained before the graphite dispersion, showing a decrease of the temperature surfaces that was related to an increase in their emissivity. The results contribute to the understanding of RCCS capability in an accident scenario.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sham, Sam [ORNL; Tan, Lizhen [ORNL; Yamamoto, Yukinori [ORNL

    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.

  14. Reliability Assessment of 2400 MWth Gas-Cooled Fast Reactor Natural Circulation Decay Heat Removal in Pressurized Situations

    Directory of Open Access Journals (Sweden)

    C. Bassi

    2008-01-01

    Full Text Available As the 2400 MWth gas-cooled fast reactor concept makes use of passive safety features in combination with active safety systems, the question of natural circulation decay heat removal (NCDHR reliability and performance assessment into the ongoing probabilistic safety assessment in support to the reactor design, named “probabilistic engineering assessment” (PEA, constitutes a challenge. Within the 5th Framework Program for Research and Development (FPRD of the European Community, a methodology has been developed to evaluate the reliability of passive systems characterized by a moving fluid and whose operation is based on physical principles, such as the natural circulation. This reliability method for passive systems (RMPSs is based on uncertainties propagation into thermal-hydraulic (T-H calculations. The aim of this exercise is finally to determine the performance reliability of the DHR system operating in a “passive” mode, taking into account the uncertainties of parameters retained for thermal-hydraulical calculations performed with the CATHARE 2 code. According to the PEA preliminary results, exhibiting the weight of pressurized scenarios (i.e., with intact primary circuit boundary for the core damage frequency (CDF, the RMPS exercise is first focusing on the NCDHR performance at these T-H conditions.

  15. Numerical Analysis on the Free Fall Motion of the Control Rod Assembly for the Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se-Hong; Choi, Choengryul; Son, Sung-Man [ELSOLTEC, Yongin (Korea, Republic of); Kim, Jae-Yong; Yoon, Kyung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    On receiving the scram signal, the control rod assemblies are released to fall into the reactor core by its weight. Thus drop time and falling velocity of the control rod assembly must be estimated for the safety evaluation. However, because of its complex shape, it is difficult to estimate the drop time by theoretical method. In this study, numerical analysis has been carried out in order to estimate drop time and falling velocity of the control rod assembly to provide the underlying data for the design optimization. Numerical analysis has been carried out to estimate the drop time and falling velocity of the control rod assembly for sodium-cooled fast reactor. Before performing the numerical analysis for the control rod assembly, sphere dropping experiment has been carried out for verification of the CFD methodology. The result of the numerical analysis for the method verification is almost same as the result of the experiment. Falling velocity and drag force increase rapidly in the beginning. And then it goes to the stable state. When the piston head of the control rod assembly is inserted into the damper, the drag force increases instantaneously and the falling velocity decreases quickly. The falling velocity is reduced about 14 % by damper. The total drop time of the control rod assembly is about 1.47s. In the next study, the experiment for the control rod assembly will be carried out, and its result is going to be compared with the CFD analysis result.

  16. Methodology for sodium fire vulnerability assessment of sodium cooled fast reactor based on the Monte-Carlo principle

    Energy Technology Data Exchange (ETDEWEB)

    Song, Wei [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China); Wu, Yuanyu [ITER Organization, Route de Vinon-sur-Verdon, 13115 Saint-Paul-lès-Durance (France); Hu, Wenjun [China Institute of Atomic Energy, P. O. Box 275(34), Beijing (China); Zuo, Jiaxu, E-mail: zuojiaxu@chinansc.cn [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China)

    2015-11-15

    Highlights: • Monte-Carlo principle coupling with fire dynamic code is adopted to perform sodium fire vulnerability assessment. • The method can be used to calculate the failure probability of sodium fire scenarios. • A calculation example and results are given to illustrate the feasibility of the methodology. • Some critical parameters and experience are shared. - Abstract: Sodium fire is a typical and distinctive hazard in sodium cooled fast reactors, which is significant for nuclear safety. In this paper, a method of sodium fire vulnerability assessment based on the Monte-Carlo principle was introduced, which could be used to calculate the probabilities of every failure mode in sodium fire scenarios. After that, the sodium fire scenario vulnerability assessment of primary cold trap room of China Experimental Fast Reactor was performed to illustrate the feasibility of the methodology. The calculation result of the example shows that the conditional failure probability of key cable is 23.6% in the sodium fire scenario which is caused by continuous sodium leakage because of the isolation device failure, but the wall temperature, the room pressure and the aerosol discharge mass are all lower than the safety limits.

  17. Development of thermal-hydraulic analysis methodology for multiple modules of water-cooled breeder blanket in fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Im, Kihak [National Fusion Research Institute, 169-148, Yuseong-gu, Daejeon 305-806 (Korea, Republic of)

    2016-02-15

    Highlights: • A methodology to simulate the K-DEMO blanket system was proposed. • The results were compared with the CFD, to verify the prediction capability of MARS. • 46 Blankets in a single sector in K-DEMO were simulated using MARS-KS. • Supervisor program was devised to handle each blanket module individually. • The calculation results showed the flow rates, pressure drops, and temperatures. - Abstract: According to the conceptual design of the fusion DEMO reactor proposed by the National Fusion Research Institute of Korea, the water-cooled breeding blanket system incorporates a total of 736 blanket modules. The heat flux and neutron wall loading to each blanket module vary along their poloidal direction, and hence, thermal analysis for at least one blanket sector is required to confirm that the temperature limitations of the materials are satisfied in all the blanket modules. The present paper proposes a methodology of thermal analysis for multiple modules of the blanket system using a nuclear reactor thermal-hydraulic analysis code, MARS-KS. In order to overcome the limitations of the code, caused by the restriction on the number of computational nodes, a supervisor program was devised, which handles each blanket module separately at first, and then corrects the flow rate, considering pressure drops that occur in each module. For a feasibility test of the proposed methodology, 46 blankets in a single sector were simulated; the calculation results of the parameters, such as mass flow, pressure drops, and temperature distribution in the multiple blanket modules showed that the multi-module analysis method can be used for efficient thermal-hydraulic analysis of the fusion DEMO reactor.

  18. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Corradin, Michael [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Anderson, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Muci, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Dominguez, A. [Texas A & M Univ., College Station, TX (United States); Tokuhiro, Akira [Univ. of Idaho, Moscow, ID (United States); Hamman, K. [Univ. of Idaho, Moscow, ID (United States)

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  19. Gas-cooled reactor commercialization study: introduction scenario and commercialization analyses for process heat applications. Final report, July 8, 1977--November 30, 1977

    Energy Technology Data Exchange (ETDEWEB)

    1977-12-01

    This report identifies and presents an introduction scenario which can lead to the operation of High Temperature Gas Cooled Reactor demonstration plants for combined process heat and electric power generation applications, and presents a commercialization analysis relevant to the organizational and management plans which could implement a development program.

  20. Thermal—hydraulic Optimization of Water—cooled Center COnductor Post for Spherical Tokamaks Reactor

    Institute of Scientific and Technical Information of China (English)

    柯严; 吴宜灿; 等

    2002-01-01

    This paper proposes a conceptual structure of segmental water-cooled Center Conductor Post(CCP) to be flexble in installment and replacement.Thermal-hydraulic optimization and sensitivity analysis of key parameters are performed based on a reference fusion transmutation system with 100MW fusion power.Numerical simulation by using a commercial code PHOENICS has been carried out to be close to the thermal-hydraulic analytical results of the CCP mid-part.

  1. Thermal-hydraulic Optimization of Water-cooled Center Conductor Post for Spherical Tokamaks Reactor

    Institute of Scientific and Technical Information of China (English)

    柯严; 吴宜灿; 黄群英; 郑善良

    2002-01-01

    This paper proposes a conceptual structure of segmental water-cooled Center Con ductor Post (CCP) to be flexible in installment and replacement. Thermal-hydraulic optimization and sensitivity analysis of key parameters are performed based on a reference fusion transmutation system with 100 MW fusion power. Numerical simulation by using a commercial code PHOEN]CS has been carried out to be close to the thermal-hydraulic analytical results of the CCP mid-part.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-02-15

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

  3. 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.

    Energy Technology Data Exchange (ETDEWEB)

    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

  4. Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai [ORNL; Wilson, Dane F [ORNL; Forsberg, Charles W [ORNL

    2007-09-01

    The Advanced High-Temperature Reactor (AHTR) is a new reactor concept that uses a liquid fluoride salt coolant and a solid high-temperature fuel. Several alternative fuel types are being considered for this reactor. One set of fuel options is the use of pin-type fuel assemblies with silicon carbide (SiC) cladding. This report provides (1) an initial viability assessment of using SiC as fuel cladding and other in-core components of the AHTR, (2) the current status of SiC technology, and (3) recommendations on the path forward. Based on the analysis of requirements, continuous SiC fiber-reinforced, chemically vapor-infiltrated SiC matrix (CVI SiC/SiC) composites are recommended as the primary option for further study on AHTR fuel cladding among various industrially available forms of SiC. Critical feasibility issues for the SiC-based AHTR fuel cladding are identified to be (1) corrosion of SiC in the candidate liquid salts, (2) high dose neutron radiation effects, (3) static fatigue failure of SiC/SiC, (4) long-term radiation effects including irradiation creep and radiation-enhanced static fatigue, and (5) fabrication technology of hermetic wall and sealing end caps. Considering the results of the issues analysis and the prospects of ongoing SiC research and development in other nuclear programs, recommendations on the path forward is provided in the order or priority as: (1) thermodynamic analysis and experimental examination of SiC corrosion in the candidate liquid salts, (2) assessment of long-term mechanical integrity issues using prototypical component sections, and (3) assessment of high dose radiation effects relevant to the anticipated operating condition.

  5. Finite element based stress analysis of graphite component in high temperature gas cooled reactor core using linear and nonlinear irradiation creep models

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov; Majumdar, Saurindranath

    2015-10-15

    Highlights: • High temperature gas cooled reactor. • Finite element based stress analysis. • H-451 graphite. • Irradiation creep model. • Graphite reflector stress analysis. - Abstract: Irradiation creep plays a major role in the structural integrity of the graphite components in high temperature gas cooled reactors. Finite element procedures combined with a suitable irradiation creep model can be used to simulate the time-integrated structural integrity of complex shapes, such as the reactor core graphite reflector and fuel bricks. In the present work a comparative study was undertaken to understand the effect of linear and nonlinear irradiation creep on results of finite element based stress analysis. Numerical results were generated through finite element simulations of a typical graphite reflector.

  6. A Management Strategy for the Heavy Water Reflector Cooling System of HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, H. S.; Park, Y. C.; Lim, S. P. (and others)

    2007-11-15

    Heavy water is used as the reflector and the moderator of the HANARO research reactor. After over 10 years operation since first criticality in 1995 there arose some operational issues related with the tritium. A task force team(TFT) has been operated for 1 year since September 2006 to study and deduce resolutions of the issues concerning the tritium and the degradation of heavy water in the HANARO reflector system. The TFT drew many recommendations on the hardware upgrade, tritium containing air control, heavy water quality management, waste management, and tritium measurement system upgrade.

  7. A History of Dosimetry for the Advanced Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    Shaw Simon

    2016-01-01

    Full Text Available This paper presents a summary of the methods used in the first ∼40 years of AGR neutron dosimetry and nuclear heating calculations, and the influence of the earlier Magnox reactor dosimetry programme. While the current state-of-the-art Monte Carlo methods are extremely powerful they still require very careful consideration of the quality of the input data, nuclear data validation and variance reduction techniques; in particular, this paper examines the difficulties in assuring the adequate convergence of calculations when Monte Carlo acceleration is applied in the presence of significant streaming paths through attenuating or scattering media.

  8. A History of Dosimetry for the Advanced Gas-cooled Reactors

    Science.gov (United States)

    Shaw, Simon; Thornton, Dean

    2016-02-01

    This paper presents a summary of the methods used in the first ˜40 years of AGR neutron dosimetry and nuclear heating calculations, and the influence of the earlier Magnox reactor dosimetry programme. While the current state-of-the-art Monte Carlo methods are extremely powerful they still require very careful consideration of the quality of the input data, nuclear data validation and variance reduction techniques; in particular, this paper examines the difficulties in assuring the adequate convergence of calculations when Monte Carlo acceleration is applied in the presence of significant streaming paths through attenuating or scattering media.

  9. Mitigation of corrosion and mass transfer in sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Latge, C. [CEA Cadarache, Dir. de l' Energie Nucleaire, 13 - Saint-Paul-lez-Durance (France); Feron, D. [CEA Saclay, Dir. de l' Energie Nucleaire, 91 - Gif-sur-Yvette (France)

    2009-07-01

    Full text of publication follows: Several coolants can be used for the development of the Fast Reactors, as sodium, gas, lead or lead-bismuth eutectic, and have been selected in the Generation IV forum. The high density energy requires a coolant with a very good thermal conductivity. Liquid sodium is such a medium which is liquid between 97.8 up to 880 C at dynamic pressure below 4 bars, and with compatible neutron-physical properties. Its viscosity is comparable to that of water and its compatibility with metallic materials is fairly satisfactory. It is however necessary to keep the conditions of operation within a range such that corrosion is limited. Several materials are suitable for use in liquid sodium reactors, among ferritic and austenitic steels and high temperature alloys with up to 32% nickel contents. The designer has however to consider the mass transfer between materials of different compositions. The exchange and transfer of non-metallic elements such as carbon or nitrogen has to be taken into account. The corrosion mechanisms of austenitic steels have been extensively studied and described in the literature: surface cleaning, austenitic dissolution, formation of a ferrite layer, steady state equilibrium and several models have been proposed: main parameters include oxygen content, sodium velocity and steel temperature. Operating experience has shown that, if there are no cladding failures, the main source of radioactivity in the primary circuit is the activated corrosion products, like {sup 54}Mn, {sup 51}Cr,..., induced by the activation of core materials which are dissolved into the sodium and mainly deposited in the coldest parts of the reactor i.e. the Intermediate Heat Exchanger (IHX) and pumps. Radio-cobalt such as {sup 60}Co are also produced and a low fraction is deposited in primary components. The corrosion rates estimated and the contamination induced by activated corrosion products observed in SFR like Phenix, JOYO, BN600, PFR, EBR2 have

  10. A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Jorge [Univ. of Utah, Salt Lake City, UT (United States)

    2013-12-01

    The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method

  11. Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mueller, Don [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-12-01

    In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.

  12. Design of an Actinide-Burning, Lead or Lead-Bismuth Cooled Reactor that Produces Low-Cost Electricity

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth; Weaver, Kevan Dean; Davis, Cliff Bybee; MIT folks

    2000-07-01

    The purpose of this Idaho National Engineering and Environmental Laboratory (INEEL) and Massachusetts Institute of Technology (MIT) University Research Consortium (URC) 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, material compatibility, plant engineering, and coolant activation. In the area of core neutronic design, the reactivity vs. burnup and discharge isotopics of both non-fertile and fertile fuels were evaluated. An innovative core for pure actinide burning that uses streaming, fertile-free fuel assemblies was studied in depth. This particular core exhibits excellent reactivity performance upon coolant voiding, even for voids that occur in the core center, and has a transuranic (TRU) destruction rate that is comparable to the proposed accelerator transmutation of waste (ATW) facility. These studies suggest that a core can be designed to achieve a long life while maintaining safety and minimizing waste. In the area of material compatibility studies, an experimental apparatus for the investigation of the flow-assisted dissolution and precipitation (corrosion) of potential fuel cladding and structural materials has been designed and built at the INEEL. The INEEL forced-convection corrosion cell consists of a small heated vessel with a shroud and gas flow system. The corrosion cell is being used to test steel that is commercially available in the United States to temperatures above 650°C. Progress in plant engineering was made for two reactor concepts, one utilizing an indirect cycle with heat exchangers and the other utilizing a direct-contact steam cycle. The evaluation of the

  13. High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Beck; L. F. Pincock

    2011-04-01

    The purpose of this report is to identify possible issues highlighted by these lessons learned that could apply to the NGNP in reducing technical risks commensurate with the current phase of design. Some of the lessons learned have been applied to the NGNP and documented in the Preconceptual Design Report. These are addressed in the background section of this document and include, for example, the decision to use TRISO fuel rather than BISO fuel used in the Peach Bottom reactor; the use of a reactor pressure vessel rather than prestressed concrete found in Fort St. Vrain; and the use of helium as a primary coolant rather than CO2. Other lessons learned, 68 in total, are documented in Sections 2 through 6 and will be applied, as appropriate, in advancing phases of design. The lessons learned are derived from both negative and positive outcomes from prior HTGR experiences. Lessons learned are grouped according to the plant, areas, systems, subsystems, and components defined in the NGNP Preconceptual Design Report, and subsequent NGNP project documents.

  14. FAST and SAFE Passive Safety Devices for Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Kim, Chihyung; Kim, In-Hyung; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The major factor is the impact of the neutron spectral hardening. The second factor that affects the CVR is reduced capture by the coolant when the coolant voiding occurs. To improve the CVR, many ideas and concepts have been proposed, which include introduction of an internal blanket, spectrum softening, or increasing the neutron leakage. These ideas may reduce the CVR, but they deteriorate the neutron economy. Another potential solution is to adopt a passive safety injection device such as the ARC (autonomous reactivity control) system, which is still under development. In this paper, two new concepts of passive safety devices are proposed. The devices are called FAST (Floating Absorber for Safety at Transient) and SAFE (Static Absorber Feedback Equipment). Their purpose is to enhance the negative reactivity feedback originating from the coolant in fast reactors. SAFE is derived to balance the positive reactivity feedback due to sodium coolant temperature increases. It has been demonstrated that SAFE allows a low-leakage SFR to achieve a self-shutdown and self-controllability even though the generic coolant temperature coefficient is quite positive and the coolant void reactivity can be largely managed by the new FAST device. It is concluded that both FAST and SAFE devices will improve substantially the fast reactor safety and they deserve more detailed investigations.

  15. Monte Carlo analysis of Very High Temperature gas-cooled Reactor for hydrogen production

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. G.; Kim, H. C.; Kim, S. Y.; Shin, C. H.; Han, C. Y.; Kim, J. C. [Hanyang Univ., Seoul (Korea, Republic of)

    2006-03-15

    This work has been pursued during 2 years. In the first year, the development of Monte Carlo analysis method for pebble-type VHTR core was focused with zero-power reactor. The pebble-bed cores of HTR-PROTEUS critical facility in Switzerland were selected for the benchmark model and detailed full-scope MCNP modeling was carried out. Especially, accurate and effective modeling of UO{sub 2} particles and their distributions in fuel pebble was pursed as well as the pebbles distribution within core region. After the detailed MCNP modeling of the whole facility, analyses of nuclear characteristics were carried out, and the results were compared with experiments and those of other research groups. The effective multiplication factors (k{sub eff}) were calculated for the two HTR-PROTEUS cores, and then homogenization effect of TRISO fuel on criticality investigated. Control rod and shutdown rod worths were also calculated, and the criticality calculations with different cross-section library and various reflector thickness were carried out. In the 2nd year of the research period, the Monte Carol analysis method developed in the 1st year was applied to the core with thermal power. The pebble-bed cores of HTR-10 test reactor in China were selected for the benchmark model. After the detailed full-scope MCNP modeling the Monte Carlo analysis results calculated in this work were verified with the benchmark results which have been done for first criticality state and initial core.

  16. CFD Analysis of Upper Plenum Flow for a Sodium-Cooled Small Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kraus, A.; Hu, R.

    2015-01-01

    Upper plenum flow behavior is important for many operational and safety issues in sodium fast reactors. The Prototype Gen-IV Sodium Fast Reactor (PGSFR), a pool-type, 150 MWe output power design, was used as a reference case for a detailed characterization of upper plenum flow for normal operating conditions. Computational Fluid Dynamics (CFD) simulation was utilized with detailed geometric modeling of major structures. Core outlet conditions based on prior system-level calculations were mapped to approximate the outlet temperatures and flow rates for each core assembly. Core outlet flow was found to largely bypass the Upper Internal Structures (UIS). Flow curves over the shield and circulates within the pool before exiting the plenum. Cross-flows and temperatures were evaluated near the core outlet, leading to a proposed height for the core outlet thermocouples to ensure accurate assembly-specific temperature readings. A passive scalar was used to evaluate fluid residence time from core outlet to IHX inlet, which can be used to assess the applicability of various methods for monitoring fuel failure. Additionally, the gas entrainment likelihood was assessed based on the CFD simulation results. Based on the evaluation of velocity gradients and turbulent kinetic energies and the available gas entrainment criteria in the literature, it was concluded that significant gas entrainment is unlikely for the current PGSFR design.

  17. Work Domain Analysis of a Predecessor Sodium-cooled Reactor as Baseline for AdvSMR Operational Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Ronald Farris; David Gertman; Jacques Hugo

    2014-03-01

    This report presents the results of the Work Domain Analysis for the Experimental Breeder Reactor (EBR-II). This is part of the phase of the research designed to incorporate Cognitive Work Analysis in the development of a framework for the formalization of an Operational Concept (OpsCon) for Advanced Small Modular Reactors (AdvSMRs). For a new AdvSMR design, information obtained through Cognitive Work Analysis, combined with human performance criteria, can and should be used in during the operational phase of a plant to assess the crew performance aspects associated with identified AdvSMR operational concepts. The main objective of this phase was to develop an analytical and descriptive framework that will help systems and human factors engineers to understand the design and operational requirements of the emerging generation of small, advanced, multi-modular reactors. Using EBR-II as a predecessor to emerging sodium-cooled reactor designs required the application of a method suitable to the structured and systematic analysis of the plant to assist in identifying key features of the work associated with it and to clarify the operational and other constraints. The analysis included the identification and description of operating scenarios that were considered characteristic of this type of nuclear power plant. This is an invaluable aspect of Operational Concept development since it typically reveals aspects of future plant configurations that will have an impact on operations. These include, for example, the effect of core design, different coolants, reactor-to-power conversion unit ratios, modular plant layout, modular versus central control rooms, plant siting, and many more. Multi-modular plants in particular are expected to have a significant impact on overall OpsCon in general, and human performance in particular. To support unconventional modes of operation, the modern control room of a multi-module plant would typically require advanced HSIs that would

  18. Development status and operational features of the high temperature gas-cooled reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Winkleblack, R.K.

    1976-04-01

    The objective of this study is to investigate the maturity of HTR-technology and to look out for possible technical problems, concerning introduction of large HTR power plants into the market. Further state and problems of introducing and closing the thorium fuel cycle is presented and judged. Finally, the state of development of advanced HTR-concepts for electricity production, the direct cycle HTR with helium turbine, and the gas-cooled fast breeder is discussed. In preparing the study, both HTR concepts with spherical and block-type fuel elements have been considered.

  19. Scientific-technical cooperation with Russia. Transient analyses for alternative types of water-cooled reactors. Final report; WTZ mit Russland. Transientenanalysen fuer wassergekuehlte Kernreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Rohde, Ulrich [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Kozmenkov, Yaroslav [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Institute of Physics and Power Engineering, Obninsk (Russian Federation); Pivovarov, Valeri; Matveev, Yurij [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    2010-12-15

    The recently developed multi-group version DYN3D-MG of the reactor dynamics code DYN3D has been qualified for applications to water-cooled reactor concepts different from industrial PWR and BWR. An extended DYN3D version was applied to the graphite-moderated pressure tube reactor EGP-6 (NPP Bilibino) and conceptual design studies of an advanced Boiling Water Reactor with reduced moderation (RMWR) as well as the RUTA-70 reactor for low temperature heat supply. Concerning the RUTA reactor, safe heat removal by natural circulation of the coolant at low pressure has to be shown. For the corresponding validation of thermo-hydraulic system codes like ATHLET and RELAP5, experiments on flashing-induced natural circulation instabilities performed at the CIRCUS test facility at the TU Delft were simulated using the RELAP5 code. For the application to alternative water-cooled reactors, DYN3D model extensions and modifications were implemented, in particular adaptations of heat conduction and heat transfer models. Performing code-to-code comparisons with the Russian fine-mesh neutron diffusion code ACADEM contributed to the verification of DYN3D-MG. Validation has been performed by calculating reactor dynamics experiments at the NPP Bilibino. For the reactors EGP-6, RMWR and RUTA, analyses of various protected and unprotected control rod withdrawal and ejection transients were performed. The beyond design basis accident (BDBA) scenario ''Coast-down of all main coolant pumps at nominal power without scram'' for the RUTA reactor was analyzed using the code complexes DYN3D/ATHLET and DYN3D/RELAP5. It was shown, that the reactor passes over to a save asymptotic state at reduced power with coolant natural circulation. Analyzing the BDBA ''Unprotected withdrawal of a control rod group'' for the RMWR, the safety against Departure from Nucleate Boiling (DNB) could not be shown with the necessary confidence. Finally, conclusions have been drawn

  20. Effects of cooling channel blockage on fuel plate temperature in Tehran Research Reactor

    Institute of Scientific and Technical Information of China (English)

    TABBAKH Farshid

    2009-01-01

    In this study, the variation of the temperature distribution of the fuel plate in Tehran Research Reactor core was studied in case of coolant channels blockage. While the experimental method is not possible, both the analytical and simulation methods were used to obtain the more reliable data. The results show that one channel blockage will increase the fuel temperature to about 100%, but it does not lead to clad melt down still. With further calculation and simulation it is understood that if the coolant velocity drops to 90% of its nominal value, it may causes the clad melt-ing down. At least two channels with complete blockage even at the positions far from the core center can also melt down the clad.

  1. Study of the reactivity effect of bubble collapse in molten liquid metal cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Munno, F.J.; Sheaks, O.J.

    1977-08-01

    The reactivity effect of void collapse on a Core Disruptive Accident in a sodium free fast reactor core was studied. A two-dimensional transport theory computer code was employed to obtain a reactivity model which incorporated random distributions of voids in spatial location and bubble size. This model was implemented into VENUS-II and several excursions were investigated for the purpose of assessing the effect of bubble collapse reactivity on excursion energy. In addition, comparisons were made between the VENUS-II results obtained using the transport reactivity model and those obtained using a reactivity model developed from the Behrens' formulation of the leakage effects due to the presence of bubbles. The results indicate that excursion energy is sensitive to the shape of the reactivity versus void fraction function and that reactivity models based strictly on leakage effects are probably conservative compared to models which include both self-multiplication and leakage.

  2. Passive acoustic leak detection for sodium cooled fast reactors using hidden Markov models

    Energy Technology Data Exchange (ETDEWEB)

    Riber Marklund, A. [CEA, Cadarache, DEN/DTN/STCP/LIET, Batiment 202, 13108 St Paul-lez-Durance, (France); Kishore, S. [Fast Reactor Technology Group of IGCAR, (India); Prakash, V. [Vibrations Diagnostics Division, Fast Reactor Technology Group of IGCAR, (India); Rajan, K.K. [Fast Reactor Technology Group and Engineering Services Group of IGCAR, (India)

    2015-07-01

    Acoustic leak detection for steam generators of sodium fast reactors have been an active research topic since the early 1970's and several methods have been tested over the years. Inspired by its success in the field of automatic speech recognition, we here apply hidden Markov models (HMM) in combination with Gaussian mixture models (GMM) to the problem. To achieve this, we propose a new feature calculation scheme, based on the temporal evolution of the power spectral density (PSD) of the signal. Using acoustic signals recorded during steam/water injection experiments done at the Indira Gandhi Centre for Atomic Research (IGCAR), the proposed method is tested. We perform parametric studies on the HMM+GMM model size and demonstrate that the proposed method a) performs well without a priori knowledge of injection noise, b) can incorporate several noise models and c) has an output distribution that simplifies false alarm rate control. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-06-01

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

  4. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Progress Report for Work Through September 2003, 2nd Annual/8th Quarterly Report

    Energy Technology Data Exchange (ETDEWEB)

    Philip E. MacDonald

    2003-09-01

    The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation-IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% vs. about 33% efficiency for current Light Water Reactors, LWRs) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus the need for recirculation and jet pumps, a pressurizer, steam generators, steam separators and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies, LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which is also in use around the world.

  5. Adaptation and implementation of the TRACE code for transient analysis in designs lead cooled fast reactors; Adaptacion y aplicacion del codigo TRACE para el analisis de transitorios en disenos de reactores rapidos refrigerados por plomo

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, A.; Ammirabile, L.; Martorell, S.

    2015-07-01

    Lead-Cooled Fast Reactor (LFR) has been identified as one of promising future reactor concepts in the technology road map of the Generation IVC International Forum (GIF)as well as in the Deployment Strategy of the European Sustainable Nuclear Industrial Initiative (ESNII), both aiming at improved sustainability, enhanced safety, economic competitiveness, and proliferation resistance. This new nuclear reactor concept requires the development of computational tools to be applied in design and safety assessments to confirm improved inherent and passive safety features of this design. One approach to this issue is to modify the current computational codes developed for the simulation of Light Water Reactors towards their applicability for the new designs. This paper reports on the performed modifications of the TRACE system code to make it applicable to LFR safety assessments. The capabilities of the modified code are demonstrated on series of benchmark exercises performed versus other safety analysis codes. (Author)

  6. Army Gas-Cooled Reactor Systems program: alternator final design report

    Energy Technology Data Exchange (ETDEWEB)

    1964-06-01

    The development and testing of a demonstration brushless alternator for the ML-1 mobile nuclear power plant is described. The brushless concept was selected after it became apparent that a conventional power generator could not satisfy the ML-1 weight and size requirements. The demonstration alternator fabricated and tested under this program did not meet all performance specifications; the efficiency was low and the unit could not be operated for significant periods of time without overheating. However, a large body of useful data was accumulated during the extensive development program. Of special interest are data on the rotor and stator design, the cooling requirements and on the distribution of eddy current losses. Analysis of the data indicates that a brushless alternator, only slightly larger and heavier than was specified for the ML-1, could be developed with a modest additional effort.

  7. Use of fine-grained shredder dust as a cement admixture after a melting, rapid-cooling and pulverizing process

    Energy Technology Data Exchange (ETDEWEB)

    Kakimoto, K. [Kyushu Institute of Technology, Fukuoka (Japan). Department of Applied Chemistry; Nakano, Y. [Kyushu Institute of Technology, Fukuoka (Japan). Graduate School, Department of Engineering; Yamasaki, T.; Shimuzu, K.; Idemitsu, T. [Kyushu Institute of Technology, Fukyoka (Japan). Department of Civil Engineering

    2004-12-01

    Shredder dust is a residue, which is removed from valuable ferrous metals found in scrap automobile and electronic waste. It is also an industrial waste byproduct which, under legislation in place since April 1996, must be disposed of in landfill sites. One method of disposing shredder dust is by scorification, however, this is a costly process and therefore impractical. Costs could be reduced if the shredder dust had a valuable use, and, in this paper, the authors examine its effectiveness as a cement admixture. First, molten shredder dust was crushed for use as a cement admixture. However, it was difficult to crush it completely because metallic grains were mixed in with molten shredder dust. These particles were removed by sifting and the molten shredder dust was crushed once again. Eventually, a fine 75 {mu}m and less powder type of slag was obtained. This slag was mixed with Ordinary Portland Cement (OPC) to form a cement mortar and subsequently a mortar test was conducted. From the test results, it was found that the long-term strength of the cement did not deteriorate even when it included 30% by weight of the pulverized molten shredder dust. (author)

  8. An Innovative Passive Residual Heat Removal System of an Open-Pool Type Research Reactor with Pump Flywheel and Gravity Core Cooling Tank

    Directory of Open Access Journals (Sweden)

    Kwon-Yeong Lee

    2015-01-01

    Full Text Available In an open-pool type research reactor, the primary cooling system can be designed to have a downward flow inside the core during normal operation because of the plate type fuel geometry. There is a flow inversion inside the core from the downward flow by the inertia force of the primary coolant to the upward flow by the natural circulation when the pump is turned off. To delay the flow inversion time, an innovative passive system with pump flywheel and GCCT is developed to remove the residual heat. Before the primary cooling pump starts up, the water level of the GCCT is the same as that of the reactor pool. During the primary cooling pump operation, the water in the GCCT is moved into the reactor pool because of the pump suction head. After the pump stops, the potential head generates a downward flow inside the core by moving the water from the reactor pool to the GCCT and removes the residual heat. When the water levels of the two pools are the same again, the core flow has an inversion of the flow direction, and natural circulation is developed through the flap valves.

  9. Integral and Separate Effects Tests for Thermal Hydraulics Code Validation for Liquid-Salt Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per

    2012-10-30

    The objective of the 3-year project was to collect integral effects test (IET) data to validate the RELAP5-3D code and other thermal hydraulics codes for use in predicting the transient thermal hydraulics response of liquid salt cooled reactor systems, including integral transient response for forced and natural circulation operation. The reference system for the project is a modular, 900-MWth Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a specific type of Fluoride salt-cooled High temperature Reactor (FHR). Two experimental facilities were developed for thermal-hydraulic integral effects tests (IETs) and separate effects tests (SETs). The facilities use simulant fluids for the liquid fluoride salts, with very little distortion to the heat transfer and fluid dynamics behavior. The CIET Test Bay facility was designed, built, and operated. IET data for steady state and transient natural circulation was collected. SET data for convective heat transfer in pebble beds and straight channel geometries was collected. The facility continues to be operational and will be used for future experiments, and for component development. The CIET 2 facility is larger in scope, and its construction and operation has a longer timeline than the duration of this grant. The design for the CIET 2 facility has drawn heavily on the experience and data collected on the CIET Test Bay, and it was completed in parallel with operation of the CIET Test Bay. CIET 2 will demonstrate start-up and shut-down transients and control logic, in addition to LOFC and LOHS transients, and buoyant shut down rod operation during transients. Design of the CIET 2 Facility is complete, and engineering drawings have been submitted to an external vendor for outsourced quality controlled construction. CIET 2 construction and operation continue under another NEUP grant. IET data from both CIET facilities is to be used for validation of system codes used for FHR modeling, such as RELAP5-3D. A set of

  10. Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors. Publishable Final Activity Report

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C., E-mail: kuijper@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Somers, J.; Van Den Durpel, L.; Chauvet, V.; Cerullo, N.; Cetnar, J.; Abram, T.; Bakker, K.; Bomboni, E.; Bernnat, W.; Domanska, J.G.; Girardi, E.; De Haas, J.B.M.; Hesketh, K.; Hiernaut, J.P.; Hossain, K.; Jonnet, J.; Kim, Y.; Kloosterman, J.L.; Kopec, M.; Murgatroyd, J.; Millington, D.; Lecarpentier, D.; Lomonaco, G.; McEachern, D.; Meier, A.; Mignanelli, M.; Nabielek, H.; Oppe, J.; Petrov, B.Y.; Pohl, C.; Ruetten, H.J.; Schihab, S.; Toury, G.; Trakas, C.; Venneri, F.; Verfondern, K.; Werner, H.; Wiss, T.; Zakova, J.

    2010-11-15

    The PUMA project -the acronym stands for 'Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors'- was a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Program (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO{sub 2}-free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR

  11. Modeling of the Reactor Core Isolation Cooling Response to Beyond Design Basis Operations - Interim Report

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Kyle [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wilson, Chisom Shawn [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Morrow, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Osborn, Douglas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gauntt, Randall O. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-12-01

    Efforts are being pursued to develop and qualify a system-level model of a reactor core isolation (RCIC) steam-turbine-driven pump. The model is being developed with the intent of employing it to inform the design of experimental configurations for full-scale RCIC testing. The model is expected to be especially valuable in sizing equipment needed in the testing. An additional intent is to use the model in understanding more fully how RCIC apparently managed to operate far removed from its design envelope in the Fukushima Daiichi Unit 2 accident. RCIC modeling is proceeding along two avenues that are expected to complement each other well. The first avenue is the continued development of the system-level RCIC model that will serve in simulating a full reactor system or full experimental configuration of which a RCIC system is part. The model reasonably represents a RCIC system today, especially given design operating conditions, but lacks specifics that are likely important in representing the off-design conditions a RCIC system might experience in an emergency situation such as a loss of all electrical power. A known specific lacking in the system model, for example, is the efficiency at which a flashing slug of water (as opposed to a concentrated jet of steam) could propel the rotating drive wheel of a RCIC turbine. To address this specific, the second avenue is being pursued wherein computational fluid dynamics (CFD) analyses of such a jet are being carried out. The results of the CFD analyses will thus complement and inform the system modeling. The system modeling will, in turn, complement the CFD analysis by providing the system information needed to impose appropriate boundary conditions on the CFD simulations. The system model will be used to inform the selection of configurations and equipment best suitable of supporting planned RCIC experimental testing. Preliminary investigations with the RCIC model indicate that liquid water ingestion by the turbine

  12. Study of a fuel assembly for the nuclear reactor of IV generation cooled with supercritical water; Estudio de un ensamble de combustible para el reactor nuclear de generacion IV enfriado con agua supercritica

    Energy Technology Data Exchange (ETDEWEB)

    Barragan M, A.; Martin del Campo M, C.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Espinosa P, G., E-mail: albrm29@yahoo.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (MX)

    2011-11-15

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