Science.gov

Sample records for reactor cycle scheme

  1. Hybrid reactors. [Fuel cycle

    SciTech Connect

    Moir, R.W.

    1980-09-09

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m/sup -2/, and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid.

  2. Recycling scheme for twin BWRs reactors

    SciTech Connect

    Ramirez-Sanchez, J. R.; Perry, R. T.; Gustavo Alonso, V.; Javier Palacios, H.

    2006-07-01

    To asses the advantages of reprocess and recycle the spent fuel from nuclear power reactors, against a once through policy, a MOX fuel design is proposed to match a generic scenario for twin BWRs and establish a fuel management scheme. Calculations for the amount of fuel that the plants will use during 40 years of operation were done, and an evaluation of costs using constant money method for each option applying current prices for uranium and services were made. Finally a comparison between the options was made, resulting that even the current high prices of uranium, still the recycling option is more expensive that the once through alternative. But reprocessing could be an alternative to reduce the amount of spent fuel stored in the reactor pools. (authors)

  3. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

    NASA Astrophysics Data System (ADS)

    Widiawati, Nina; Su'ud, Zaki

    2015-09-01

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from -0.6695443 % at BOC to -0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

  4. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

    SciTech Connect

    Widiawati, Nina Su’ud, Zaki

    2015-09-30

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from −0.6695443 % at BOC to −0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

  5. Reactivity Control Schemes for Fast Spectrum Space Nuclear Reactors

    SciTech Connect

    Craft, Aaron E.; King, Jeffrey C.

    2008-01-21

    Several different reactivity control schemes are considered for future space nuclear reactor power systems. Each of these control schemes uses a combination of boron carbide absorbers and/or beryllium oxide reflectors to achieve sufficient reactivity swing to keep the reactor subcritical during launch and to provide sufficient excess reactivity to operate the reactor over its expected 7-15 year lifetime. The size and shape of the control system directly impacts the size and mass of the space reactor's reflector and shadow shield, leading to a tradeoff between reactivity swing and total system mass. This paper presents a trade study of drum, shutter, and petal control schemes based on reactivity swing and mass effects for a representative fast-spectrum, gas-cooled reactor. For each control scheme, the dimensions and composition of the core are constant, and the reflector is sized to provide $5 of cold-clean excess reactivity with each configuration in its most reactive state. The advantages and disadvantages of each configuration are discussed, along with optimization techniques and novel geometric approaches for each scheme.

  6. Coupling Schemes for Multiphysics Reactor Simulation

    SciTech Connect

    Vijay Mahadeven; Jean Ragusa

    2007-11-01

    This report documents the progress of the student Vijay S. Mahadevan from the Nuclear Engineering Department of Texas A&M University over the summer of 2007 during his visit to the INL. The purpose of his visit was to investigate the physics-based preconditioned Jacobian-free Newton-Krylov method applied to physics relevant to nuclear reactor simulation. To this end he studied two test problems that represented reaction-diffusion and advection-reaction. These two test problems will provide the basis for future work in which neutron diffusion, nonlinear heat conduction, and a twophase flow model will be tightly coupled to provide an accurate model of a BWR core.

  7. Steady-State Analysis Model for Advanced Fuel Cycle Schemes.

    Energy Science and Technology Software Center (ESTSC)

    2008-03-17

    Version 00 SMAFS was developed as a part of the study, "Advanced Fuel Cycles and Waste Management", which was performed during 2003-2005 by an ad-hoc expert group under the Nuclear Development Committee in the OECD/NEA. The model was designed for an efficient conduct of nuclear fuel cycle scheme cost analyses. It is simple, transparent and offers users the capability to track down cost analysis results. All the fuel cycle schemes considered in the model aremore » represented in a graphic format and all values related to a fuel cycle step are shown in the graphic interface, i.e., there are no hidden values embedded in the calculations. All data on the fuel cycle schemes considered in the study including mass flows, waste generation, cost data, and other data such as activities, decay heat and neutron sources of spent fuel and high-level waste along time are included in the model and can be displayed. The user can easily modify values of mass flows and/or cost parameters and see corresponding changes in the results. The model calculates: front-end fuel cycle mass flows such as requirements of enrichment and conversion services and natural uranium; mass of waste based on the waste generation parameters and the mass flow; and all costs.« less

  8. Nuclear reactors and the nuclear fuel cycle

    SciTech Connect

    Pearlman, H

    1989-11-01

    According to the author, the first sustained nuclear fission chain reaction was not at the University of Chicago, but at the Oklo site in the African country of Gabon. Proof of this phenomenon is provided by mass spectrometric and analytical chemical measurements by French scientists. The U.S. experience in developing power-producing reactors and their related fuel and fuel cycles is discussed.

  9. The dynamo basis of solar cycle precursor schemes

    NASA Astrophysics Data System (ADS)

    Charbonneau, Paul; Barlet, Guillaume

    2011-02-01

    We investigate the dynamo underpinning of solar cycle precursor schemes based on direct or indirect measures of the solar surface magnetic field. We do so for various types of mean-field-like kinematic axisymmetric dynamo models, where amplitude fluctuations are driven by zero-mean stochastic forcing of the dynamo number controlling the strength of the poloidal source term. In all stochastically forced models considered, the surface poloidal magnetic field is found to have precursor value only if it feeds back into the dynamo loop, which suggests that accurate determination of the magnetic flux budget of the solar polar fields may hold the key to dynamo model-based cycle forecasting.

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

    SciTech Connect

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

    2010-06-22

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

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

    NASA Astrophysics Data System (ADS)

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

    2010-06-01

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

  12. Immobilization of Fast Reactor First Cycle Raffinate

    SciTech Connect

    Langley, K. F.; Partridge, B. A.; Wise, M.

    2003-02-26

    This paper describes the results of work to bring forward the timing for the immobilization of first cycle raffinate from reprocessing fuel from the Dounreay Prototype Fast Reactor (PFR). First cycle raffinate is the liquor which contains > 99% of the fission products separated from spent fuel during reprocessing. Approximately 203 m3 of raffinate from the reprocessing of PFR fuel is held in four tanks at the UKAEA's site at Dounreay, Scotland. Two methods of immobilization of this high level waste (HLW) have been considered: vitrification and cementation. Vitrification is the standard industry practice for the immobilization of first cycle raffinate, and many papers have been presented on this technique elsewhere. However, cementation is potentially feasible for immobilizing first cycle raffinate because the heat output is an order of magnitude lower than typical HLW from commercial reprocessing operations such as that at the Sellafield site in Cumbria, England. In fact, it falls within the upper end of the UK definition of intermediate level waste (ILW). Although the decision on which immobilization technique will be employed has yet to be made, initial development work has been undertaken to identify a suitable cementation formulation using inactive simulant of the raffinate. An approach has been made to the waste disposal company Nirex to consider the disposability of the cemented product material. The paper concentrates on the process development work that is being undertaken on cementation to inform the decision making process for selection of the immobilization method.

  13. Multi-cycle boiling water reactor fuel cycle optimization

    SciTech Connect

    Ottinger, K.; Maldonado, G.I.

    2013-07-01

    In this work a new computer code, BWROPT (Boiling Water Reactor Optimization), is presented. BWROPT uses the Parallel Simulated Annealing (PSA) algorithm to solve the out-of-core optimization problem coupled with an in-core optimization that determines the optimum fuel loading pattern. However it uses a Haling power profile for the depletion instead of optimizing the operating strategy. The result of this optimization is the optimum new fuel inventory and the core loading pattern for the first cycle considered in the optimization. Several changes were made to the optimization algorithm with respect to other nuclear fuel cycle optimization codes that use PSA. Instead of using constant sampling probabilities for the solution perturbation types throughout the optimization as is usually done in PSA optimizations the sampling probabilities are varied to get a better solution and/or decrease runtime. The new fuel types available for use can be sorted into an array based on any number of parameters so that each parameter can be incremented or decremented, which allows for more precise fuel type selection compared to random sampling. Also, the results are sorted by the new fuel inventory of the first cycle for ease of comparing alternative solutions. (authors)

  14. Impact of Geoengineering Schemes on the Global Hydrological Cycle

    SciTech Connect

    Bala, G; Duffy, P; Taylor, K

    2007-12-07

    The rapidly rising CO{sub 2} level in the atmosphere has led to proposals of climate stabilization via 'Geoengineering' schemes that would mitigate climate change by intentionally reducing the solar radiation incident on earth's surface. In this paper, we address the impact of these climate stabilization schemes on the global hydrological cycle, using equilibrium simulations from an atmospheric general circulation model coupled to a slab ocean model. We show that insolation reductions sufficient to offset global-scale temperature increases lead to a decrease in the intensity of the global hydrologic cycle. This occurs because solar forcing is more effective in driving changes in global mean evaporation than is CO{sub 2} forcing of a similar magnitude. In the model used here, the hydrologic sensitivity, defined as the percentage change in global mean precipitation per degree warming, is 2.4% for solar forcing, but only 1.5% for CO{sub 2} forcing. Although other models and the climate system itself may differ quantitatively from this result, the conclusion can be understood based on simple considerations of the surface energy budget and thus is likely to be robust. Compared to changing temperature by altering greenhouse gas concentrations, changing temperature by varying insolation results in larger changes in net radiative fluxes at the surface; these are compensated by larger changes in latent and sensible heat fluxes. Hence the hydrological cycle is more sensitive to temperature adjustment via changes in insolation than changes in greenhouse gases. This implies that an alteration in solar forcing might offset temperature changes or hydrological changes from greenhouse warming, but could not cancel both at once.

  15. Impact of geoengineering schemes on the global hydrological cycle.

    PubMed

    Bala, G; Duffy, P B; Taylor, K E

    2008-06-01

    The rapidly rising CO(2) level in the atmosphere has led to proposals of climate stabilization by "geoengineering" schemes that would mitigate climate change by intentionally reducing solar radiation incident on Earth's surface. In this article we address the impact of these climate stabilization schemes on the global hydrological cycle. By using equilibrium climate simulations, we show that insolation reductions sufficient to offset global-scale temperature increases lead to a decrease in global mean precipitation. This occurs because solar forcing is more effective in driving changes in global mean evaporation than is CO(2) forcing of a similar magnitude. In the model used here, the hydrological sensitivity, defined as the percentage change in global mean precipitation per degree warming, is 2.4% K(-1) for solar forcing, but only 1.5% K(-1) for CO(2) forcing. Although other models and the climate system itself may differ quantitatively from this result, the conclusion can be understood based on simple considerations of the surface energy budget and thus is likely to be robust. For the same surface temperature change, insolation changes result in relatively larger changes in net radiative fluxes at the surface; these are compensated by larger changes in the sum of latent and sensible heat fluxes. Hence, the hydrological cycle is more sensitive to temperature adjustment by changes in insolation than by changes in greenhouse gases. This implies that an alteration in solar forcing might offset temperature changes or hydrological changes from greenhouse warming, but could not cancel both at once. PMID:18505844

  16. Impact of geoengineering schemes on the global hydrological cycle

    PubMed Central

    Bala, G.; Duffy, P. B.; Taylor, K. E.

    2008-01-01

    The rapidly rising CO2 level in the atmosphere has led to proposals of climate stabilization by “geoengineering” schemes that would mitigate climate change by intentionally reducing solar radiation incident on Earth's surface. In this article we address the impact of these climate stabilization schemes on the global hydrological cycle. By using equilibrium climate simulations, we show that insolation reductions sufficient to offset global-scale temperature increases lead to a decrease in global mean precipitation. This occurs because solar forcing is more effective in driving changes in global mean evaporation than is CO2 forcing of a similar magnitude. In the model used here, the hydrological sensitivity, defined as the percentage change in global mean precipitation per degree warming, is 2.4% K−1 for solar forcing, but only 1.5% K−1 for CO2 forcing. Although other models and the climate system itself may differ quantitatively from this result, the conclusion can be understood based on simple considerations of the surface energy budget and thus is likely to be robust. For the same surface temperature change, insolation changes result in relatively larger changes in net radiative fluxes at the surface; these are compensated by larger changes in the sum of latent and sensible heat fluxes. Hence, the hydrological cycle is more sensitive to temperature adjustment by changes in insolation than by changes in greenhouse gases. This implies that an alteration in solar forcing might offset temperature changes or hydrological changes from greenhouse warming, but could not cancel both at once. PMID:18505844

  17. Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

    SciTech Connect

    Worrall, Andrew; Todosow, Michael

    2016-01-01

    Small modular reactors may offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of small modular reactors on the nuclear fuel cycle and fuel cycle performance. The focus of this paper is on the fuel cycle impacts of light water small modular reactors in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy Office of Nuclear Energy Fuel Cycle Options Campaign. Challenges with small modular reactors include: increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burn-up in the reactor and the fuel cycle performance. This paper summarizes the results of an expert elicitation focused on developing a list of the factors relevant to small modular reactor fuel, core, and operation that will impact fuel cycle performance. Preliminary scoping analyses were performed using a regulatory-grade reactor core simulator. The hypothetical light water small modular reactor considered in these preliminary scoping studies is a cartridge type one-batch core with 4.9% enrichment. Some core parameters, such as the size of the reactor and general assembly layout, are similar to an example small modular reactor concept from industry. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burn-up of the reactor. Fuel cycle performance

  18. Multiple reheat helium Brayton cycles for sodium fast reactors

    SciTech Connect

    Haihua Zhao; Per F. Peterson

    2008-07-01

    Sodium fast reactors (SFR) traditionally adopt the steam Rankine cycle for power conversion. The resulting potential for water-sodium reaction remains a continuing concern which at least partly delays the SFR technology commercialization and is a contributor to higher capital cost. Supercritical CO2 provides an alternative, but is also capable of sustaining energetic chemical reactions with sodium. Recent development on advanced inert-gas Brayton cycles could potentially solve this compatibility issue, increase thermal efficiency, and bring down the capital cost close to light water reactors. In this paper, helium Brayton cycles with multiple reheat and intercooling states are presented for SFRs with reactor outlet temperatures in the range of 510°C to 650°C. The resulting thermal efficiencies range from 39% and 47%, which is comparable with supercritical recompression CO2 cycles (SCO2 cycle). A systematic comparison between multiple reheat helium Brayton cycle and the SCO2 cycle is given, considering compatibility issues, plant site cooling temperature effect on plant efficiency, full plant cost optimization, and other important factors. The study indicates that the multiple reheat helium cycle is the preferred choice over SCO2 cycle for sodium fast reactors.

  19. Strengthening the nuclear-reactor fuel cycle against proliferation

    SciTech Connect

    Travelli, A.; Snelgrove, J.; Persiani, P.

    1992-12-31

    Argonne National Laboratory (ANL) conducts several research programs that serve to reduce the risks of fissile-material diversion from the nuclear-reactor fuel cycle. The objectives are to provide economical and efficient neutron or power generation with the minimum of inherent risks, and to further minimize risks by utilizing sophisticated techniques to detect attempts at material diversion. This paper will discuss the Reduced Enrichment Research and Test Reactor (RERTR) Program, the Isotope Correlation Technique (ICT), and Proliferation-Resistant Closed-Cycle Reactors. The first two are sponsored by the DOE Office of Arms Control and Nonproliferation.

  20. Waste disposal from the light water reactor fuel cycle

    NASA Astrophysics Data System (ADS)

    Costello, J. M.; Hardy, C. J.

    1981-05-01

    Alternaive nuclear fuel cycles for support of light water reactors are described and wastes containing naturally occurring or artificially produced radioactivity reviewed. General principles and objectives in radioactive waste management are outlined, and methods for their practical application to fuel cycle wastes discussed. Management of wastes from upgrading processes of uranium hexafluoride manufacture and uranium manufacture and enrichment, and, to a lesser extent, nuclear power reactor wastes are discussed. Some estimates of radiological dose commitments and health effects from nuclear power and fuel cycle wastes were made for US conditions. These indicate that the major part of the radiological dose arises from uranium mining and milling, operation of nuclear reactors, and spent fuel reprocessing. However, the total dose from the fuel cycle is estimated to be only a small fraction of that from natural background radiation.

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

    SciTech Connect

    Su'ud, Zaki; Sekimoto, H.

    2014-09-30

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

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

    NASA Astrophysics Data System (ADS)

    Su'ud, Zaki; Sekimoto, H.

    2014-09-01

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

  3. Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

    SciTech Connect

    Hardie, R.W.; Barrett, R.J.; Freiwald, J.G.

    1980-06-01

    The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and /sup 233/U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles.

  4. Proposed fuel cycle for the Integral Fast Reactor

    SciTech Connect

    Burris, L.; Walters, L.C.

    1985-01-01

    One of the key features of ANL's Integral Fast Reactor (IFR) concept is a close-coupled fuel cycle. The proposed fuel cycle is similar to that demonstrated over the first five to six years of operation of EBR-II, when a fuel cycle facility adjacent to EBR-II was operated to reprocess and refabricate rapidly fuel discharged from the EBR-II. Locating the IFR and its fuel cycle facility on the same site makes the IFR a self-contained system. Because the reactor fuel and the uranium blanket are metals, pyrometallurgical processes (shortned to ''pyroprocesses'') have been chosen. The objectives of the IFR processes for the reactor fuel and blanket materials are to (1) recover fissionable materials in high yield; (2) remove fission products adequately from the reactor fuel, e.g., a decontamination factor of 10 to 100; and (3) upgrade the concentration of plutonium in uranium sufficiently to replenish the fissile-material content of the reactor fuel. After the fuel has been reconstituted, new fuel elements will be fabricated for recycle to the reactor.

  5. Code System for Reactor Physics and Fuel Cycle Simulation.

    Energy Science and Technology Software Center (ESTSC)

    1999-04-21

    Version 00 VSOP94 (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shut-down features, in-core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterativemore » processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. In addition to its use in research and development work for the High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors.« less

  6. Code System for Reactor Physics and Fuel Cycle Simulation.

    SciTech Connect

    TEUCHERT, E.

    1999-04-21

    Version 00 VSOP94 (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shut-down features, in-core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. In addition to its use in research and development work for the High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors.

  7. An analysis of hydrogen production via closed-cycle schemes. [thermochemical processings from water

    NASA Technical Reports Server (NTRS)

    Chao, R. E.; Cox, K. E.

    1975-01-01

    A thermodynamic analysis and state-of-the-art review of three basic schemes for production of hydrogen from water: electrolysis, thermal water-splitting, and multi-step thermochemical closed cycles is presented. Criteria for work-saving thermochemical closed-cycle processes are established, and several schemes are reviewed in light of such criteria. An economic analysis is also presented in the context of energy costs.

  8. Non-Proliferative, Thorium-Based, Core and Fuel Cycle for Pressurized Water Reactors

    SciTech Connect

    Todosow M.; Todosow M.; Raitses, G. Galperin, A.

    2009-07-12

    Two of the major barriers to the expansion of worldwide adoption of nuclear power are related to proliferation potential of the nuclear fuel cycle and issues associated with the final disposal of spent fuel. The Radkowsky Thorium Fuel (RTF) concept proposed by Professor A. Radkowsky offers a partial solution to these problems. The main idea of the concept is the utilization of the seed-blanket unit (SBU) fuel assembly geometry which is a direct replacement for a 'conventional' assembly in either a Russian pressurized water reactor (VVER-1000) or a Western pressurized water reactor (PWR). The seed-blanket fuel assembly consists of a fissile (U) zone, known as seed, and a fertile (Th) zone known as blanket. The separation of fissile and fertile allows separate fuel management schemes for the thorium part of the fuel (a subcritical 'blanket') and the 'driving' part of the core (a supercritical 'seed'). The design objective for the blanket is an efficient generation and in-situ fissioning of the U233 isotope, while the design objective for the seed is to supply neutrons to the blanket in a most economic way, i.e. with minimal investment of natural uranium. The introduction of thorium as a fertile component in the nuclear fuel cycle significantly reduces the quantity of plutonium production and modifies its isotopic composition, reducing the overall proliferation potential of the fuel cycle. Thorium based spent fuel also contains fewer higher actinides, hence reducing the long-term radioactivity of the spent fuel. The analyses show that the RTF core can satisfy the requirements of fuel cycle length, and the safety margins of conventional pressurized water reactors. The coefficients of reactivity are comparable to currently operating VVER's/PWR's. The major feature of the RTF cycle is related to the total amount of spent fuel discharged for each cycle from the reactor core. The fuel management scheme adopted for RTF core designs allows a significant decrease in the

  9. Computational fluid dynamics modelling of biomass fast pyrolysis in fluidised bed reactors, focusing different kinetic schemes.

    PubMed

    Ranganathan, Panneerselvam; Gu, Sai

    2016-08-01

    The present work concerns with CFD modelling of biomass fast pyrolysis in a fluidised bed reactor. Initially, a study was conducted to understand the hydrodynamics of the fluidised bed reactor by investigating the particle density and size, and gas velocity effect. With the basic understanding of hydrodynamics, the study was further extended to investigate the different kinetic schemes for biomass fast pyrolysis process. The Eulerian-Eulerian approach was used to model the complex multiphase flows in the reactor. The yield of the products from the simulation was compared with the experimental data. A good comparison was obtained between the literature results and CFD simulation. It is also found that CFD prediction with the advanced kinetic scheme is better when compared to other schemes. With the confidence obtained from the CFD models, a parametric study was carried out to study the effect of biomass particle type and size and temperature on the yield of the products. PMID:26927234

  10. Brayton Cycle for High-Temperature Gas-Cooled Reactors

    SciTech Connect

    Oh, Chang H.; Moore, Richard L.

    2005-03-15

    This paper describes research on improving the Brayton cycle efficiency for a high-temperature gas-cooled reactor (HTGR). In this study, we are investigating the efficiency of an indirect helium Brayton cycle for the power conversion side of an HTGR power plant. A reference case based on a 250-MW(thermal) pebble bed HTGR was developed using helium gas as a working fluid in both the primary and power conversion sides. The commercial computer code HYSYS was used for process optimization. A numerical model using the Visual-Basic (V-B) computer language was also developed to assist in the evaluation of the Brayton cycle efficiency. Results from both the HYSYS simulation and the V-B model were compared with Japanese calculations based on the 300-MW(electric) Gas Turbine High-Temperature Reactor (GTHTR) that was developed by the Japan Atomic Energy Research Institute. After benchmarking our models, parametric investigations were performed to see the effect of important parameters on the cycle efficiency. We also investigated single-shaft versus multiple-shaft arrangements for the turbomachinery. The results from this study are applicable to other reactor concepts such as fast gas-cooled reactors, supercritical water reactors, and others.The ultimate goal of this study is to use other fluids such as supercritical carbon dioxide for the HTGR power conversion loop in order to improve the cycle efficiency over that of the helium Brayton cycle. This study is in progress, and the results will be published in a subsequent paper.

  11. Brayton Cycle for High Temperature Gas-Cooled Reactors

    SciTech Connect

    Chang Oh

    2005-03-01

    This paper describes research on improving the Brayton cycle efficiency for a high-temperature gas-cooled reactor (HTGR). In this study, we are investigating the efficiency of an indirect helium Brayton cycle for the power conversion side of an HTGR power plant. A reference case based on a 250-MW(thermal) pebble bed HTGR was developed using helium gas as a working fluid in both the primary and power conversion sides. The commercial computer code HYSYS was used for process optimization. A numerical model using the Visual-Basic (V-B) computer language was also developed to assist in the evaluation of the Brayton cycle efficiency. Results from both the HYSYS simulation and the V-B model were compared with Japanese calculations based on the 300-MW(electric) Gas Turbine High-Temperature Reactor (GTHTR) that was developed by the Japan Atomic Energy Research Institute. After benchmarking our models, parametric investigations were performed to see the effect of important parameters on the cycle efficiency. We also investigated single-shaft versus multiple-shaft arrangements for the turbomachinery. The results from this study are applicable to other reactor concepts such as fast gas-cooled reactors, supercritical water reactors, and others. The ultimate goal of this study is to use other fluids such as supercritical carbon dioxide for the HTGR power conversion loop in order to improve the cycle efficiency over that of the helium Brayton cycle. This study is in progress, and the results will be published in a subsequent paper.

  12. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    SciTech Connect

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  13. Supercritical Water Reactor Cycle for Medium Power Applications

    SciTech Connect

    BD Middleton; J Buongiorno

    2007-04-25

    Scoping studies for a power conversion system based on a direct-cycle supercritical water reactor have been conducted. The electric power range of interest is 5-30 MWe with a design point of 20 MWe. The overall design objective is to develop a system that has minimized physical size and performs satisfactorily over a broad range of operating conditions. The design constraints are as follows: Net cycle thermal efficiency {ge}20%; Steam turbine outlet quality {ge}90%; and Pumping power {le}2500 kW (at nominal conditions). Three basic cycle configurations were analyzed. Listed in order of increased plant complexity, they are: (1) Simple supercritical Rankine cycle; (2) All-supercritical Brayton cycle; and (3) Supercritical Rankine cycle with feedwater preheating. The sensitivity of these three configurations to various parameters, such as reactor exit temperature, reactor pressure, condenser pressure, etc., was assessed. The Thermoflex software package was used for this task. The results are as follows: (a) The simple supercritical Rankine cycle offers the greatest hardware simplification, but its high reactor temperature rise and reactor outlet temperature may pose serious problems from the viewpoint of thermal stresses, stability and materials in the core. (b) The all-supercritical Brayton cycle is not a contender, due to its poor thermal efficiency. (c) The supercritical Rankine cycle with feedwater preheating affords acceptable thermal efficiency with lower reactor temperature rise and outlet temperature. (d) The use of a moisture separator improves the performance of the supercritical Rankine cycle with feedwater preheating and allows for a further reduction of the reactor outlet temperature, thus it was selected for the next step. Preliminary engineering design of the supercritical Rankine cycle with feedwater preheating and moisture separation was performed. All major components including the turbine, feedwater heater, feedwater pump, condenser, condenser pump

  14. Thermonuclear inverse magnetic pumping power cycle for stellarator reactor

    DOEpatents

    Ho, Darwin D.; Kulsrud, Russell M.

    1991-01-01

    The plasma column in a stellarator is compressed and expanded alternatively in minor radius. First a plasma in thermal balance is compressed adiabatically. The volume of the compressed plasma is maintained until the plasma reaches a new thermal equilibrium. The plasma is then expanded to its original volume. As a result of the way a stellarator works, the plasma pressure during compression is less than the corresponding pressure during expansion. Therefore, negative work is done on the plasma over a complete cycle. This work manifests itself as a back-voltage in the toroidal field coils. Direct electrical energy is obtained from this voltage. Alternatively, after the compression step, the plasma can be expanded at constant pressure. The cycle can be made self-sustaining by operating a system of two stellarator reactors in tandem. Part of the energy derived from the expansion phase of a first stellarator reactor is used to compress the plasma in a second stellarator reactor.

  15. Energetic closed-cycle gas core reactors for orbit raising

    NASA Technical Reports Server (NTRS)

    Rosa, R. J.; Myrabo, L. N.

    1983-01-01

    Closed-cycle gas core reactor power plants can be of two types. In the 'mixed flow' type, the gaseous nuclear fuel is intimately mixed with the working gas in the cavity. In the 'light bulb' type the fissioning plasma is enclosed in a transparent tube, and energy transfer to the separate working gas occurs by thermal radiation. The potentials of high temperature gas core reactors in terrestrial electric power generator applications have been considered, and a number of civilian power-beaming applications for gaseous fuel nuclear-MHD power plants in space have been suggested. Major conclusions of investigations related to the design of space power systems are discussed. Attention is given to options for conversion cycles, the power system specific mass, and research and technology issues.

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

    NASA Astrophysics Data System (ADS)

    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

  17. Closed Brayton cycle power conversion systems for nuclear reactors :

    SciTech Connect

    Wright, Steven A.; Lipinski, Ronald J.; Vernon, Milton E.; Sanchez, Travis

    2006-04-01

    This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at

  18. Leap Frog and Time Step Sub-Cycle Scheme for Coupled Neutronics and Thermal-Hydraulic Codes

    SciTech Connect

    Lu, S.

    2002-07-01

    As the result of the advancing TCP/IP based inter-process communication technology, more and more legacy thermal-hydraulic codes have been coupled with neutronics codes to provide best-estimate capabilities for reactivity related reactor transient analysis. Most of the coupling schemes are based on closely coupled serial or parallel approaches. Therefore, the execution of the coupled codes usually requires significant CPU time, when a complicated system is analyzed. Leap Frog scheme has been used to reduce the run time. The extent of the decoupling is usually determined based on a trial and error process for a specific analysis. It is the intent of this paper to develop a set of general criteria, which can be used to invoke the automatic Leap Frog algorithm. The algorithm will not only provide the run time reduction but also preserve the accuracy. The criteria will also serve as the base of an automatic time step sub-cycle scheme when a sudden reactivity change is introduced and the thermal-hydraulic code is marching with a relatively large time step. (authors)

  19. DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

    DOEpatents

    Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

    1962-08-14

    A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

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

    SciTech Connect

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

    2008-04-01

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

  1. A Supercritical CO{sub 2} Gas Turbine Power Cycle for Next-Generation Nuclear Reactors

    SciTech Connect

    Dostal, Vaclav; Driscoll, Michael J.; Hejzlar, Pavel; Todreas, Neil E.

    2002-07-01

    Although proposed more than 35 years ago, the use of supercritical CO{sub 2} as the working fluid in a closed circuit Brayton cycle has so far not been implemented in practice. Industrial experience in several other relevant applications has improved prospects, and its good efficiency at modest temperatures (e.g., {approx}45% at 550 deg. C) make this cycle attractive for a variety of advanced nuclear reactor concepts. The version described here is for a gas-cooled, modular fast reactor. In the proposed gas-cooled fast breeder reactor design of present interest, CO{sub 2} is also especially attractive because it allows the use of metal fuel and core structures. The principal advantage of a supercritical CO{sub 2} Brayton cycle is its reduced compression work compared to an ideal gas such as helium: about 15% of gross power turbine output vs. 40% or so. This also permits the simplification of use of a single compressor stage without inter-cooling. The requisite high pressure ({approx}20 MPa) also has the benefit of more compact heat exchangers and turbines. Finally, CO{sub 2} requires significantly fewer turbine stages than He, its principal competitor for nuclear gas turbine service. One disadvantage of CO{sub 2} in a direct cycle application is the production of N-16, which will require turbine plant shielding (albeit much less than in a BWR). The cycle efficiency is also very sensitive to recuperator effectiveness and compressor inlet temperature. It was found necessary to split the recuperator into separate high-and low-temperature components, and to employ intermediate re-compression, to avoid having a pinch-point in the cold end of the recuperator. Over the past several decades developments have taken place that make the acceptance of supercritical CO{sub 2} systems more likely: supercritical CO{sub 2} pipelines are in use in the western US in oil-recovery operations; 14 advanced gas-cooled reactors (AGR) are employed in the UK at CO{sub 2} temperatures up to

  2. Synfuels from fusion: producing hydrogen with the Tandem Mirror Reactor and thermochemical cycles

    SciTech Connect

    Werner, R.W.; Ribe, F.L.

    1981-01-21

    This volume contains the following sections: (1) the Tandem Mirror fusion driver, (2) the Cauldron blanket module, (3) the flowing microsphere, (4) coupling the reactor to the process, (5) the thermochemical cycles, and (6) chemical reactors and process units. (MOW)

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

    NASA Astrophysics Data System (ADS)

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

    2010-06-01

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

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

    SciTech Connect

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

    2010-06-22

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

  5. Supercritical CO2 direct cycle Gas Fast Reactor (SC-GFR) concept.

    SciTech Connect

    Wright, Steven Alan; Parma, Edward J., Jr.; Suo-Anttila, Ahti Jorma; Al Rashdan, Ahmad; Tsvetkov, Pavel Valeryevich; Vernon, Milton E.; Fleming, Darryn D.; Rochau, Gary Eugene

    2011-05-01

    This report describes the supercritical carbon dioxide (S-CO{sub 2}) direct cycle gas fast reactor (SC-GFR) concept. The SC-GFR reactor concept was developed to determine the feasibility of a right size reactor (RSR) type concept using S-CO{sub 2} as the working fluid in a direct cycle fast reactor. Scoping analyses were performed for a 200 to 400 MWth reactor and an S-CO{sub 2} Brayton cycle. Although a significant amount of work is still required, this type of reactor concept maintains some potentially significant advantages over ideal gas-cooled systems and liquid metal-cooled systems. The analyses presented in this report show that a relatively small long-life reactor core could be developed that maintains decay heat removal by natural circulation. The concept is based largely on the Advanced Gas Reactor (AGR) commercial power plants operated in the United Kingdom and other GFR concepts.

  6. The IAEA international conference on fast reactors and related fuel cycles: highlights and main outcomes

    SciTech Connect

    Monti, S.; Toti, A.

    2013-07-01

    The 'International Conference on Fast Reactors and Related Fuel Cycles', which is regularly held every four years, represents the main international event dealing with fast reactors technology and related fuel cycles options. Main topics of the conference were new fast reactor concepts, design and simulation capabilities, safety of fast reactors, fast reactor fuels and innovative fuel cycles, analysis of past experience, fast reactor knowledge management. Particular emphasis was put on safety aspects, considering the current need of developing and harmonizing safety standards for fast reactors at the international level, taking also into account the lessons learned from the accident occurred at the Fukushima- Daiichi nuclear power plant in March 2011. Main advances in the several key areas of technological development were presented through 208 oral presentations during 41 technical sessions which shows the importance taken by fast reactors in the future of nuclear energy.

  7. METHOD AND APPARATUS FOR CONTROLLING DIRECT-CYCLE NEUTRONIC REACTORS

    DOEpatents

    Reed, G.A.

    1961-01-10

    A control arrangement is offered for a boiling-water reactor. Boric acid is maintained in the water in the reactor and the amount in the reactor is controlled by continuously removing a portion of the water from the reactor, concentrating the boric acid by evaporating the water therefrom, returning a controlled amount of the acid to the reactor, and simultaneously controlling the water level by varying the rate of spent steam return to the reactor.

  8. Numerical simulation of transonic limit cycle oscillations using high-order low-diffusion schemes

    NASA Astrophysics Data System (ADS)

    Wang, Baoyuan; Zha, Ge-Cheng

    2010-05-01

    This paper simulates the NLR7301 airfoil limit cycle oscillation (LCO) caused by fluid-structure interaction (FSI) using Reynolds averaged Navier-Stokes equations (RANS) coupled with Spalart-Allmaras (S-A) one-equation turbulence model. A low diffusion E-CUSP (LDE) scheme with 5th order weighted essentially nonoscillatory scheme (WENO) is employed to calculate the inviscid fluxes. A fully conservative 4th order central differencing is used for the viscous terms. A fully coupled fluid-structural interaction model is employed. For the case computed in this paper, the predicted LCO frequency, amplitudes, averaged lift and moment, all agree excellently with the experiment performed by Schewe et al. The solutions appear to have bifurcation and are dependent on the initial fields or initial perturbation. The developed computational fluid dynamics (CFD)/computational structure dynamics (CSD) simulation is able to capture the LCO with very small amplitudes measured in the experiment. This is attributed to the high order low diffusion schemes, fully coupled FSI model, and the turbulence model used. This research appears to be the first time that a numerical simulation of LCO matches the experiment. The simulation confirms several observations of the experiment.

  9. Dynamo-based scheme for forecasting the magnitude of solar activity cycles

    NASA Technical Reports Server (NTRS)

    Layden, A. C.; Fox, P. A.; Howard, J. M.; Sarajedini, A.; Schatten, K. H.

    1991-01-01

    This paper presents a general framework for forecasting the smoothed maximum level of solar activity in a given cycle, based on a simple understanding of the solar dynamo. This type of forecasting requires knowledge of the sun's polar magnetic field strength at the preceding activity minimum. Because direct measurements of this quantity are difficult to obtain, the quality of a number of proxy indicators already used by other authors is evaluated, which are physically related to the sun's polar field. These indicators are subjected to a rigorous statistical analysis, and the analysis technique for each indicator is specified in detail in order to simplify and systematize reanalysis for future use. It is found that several of these proxies are in fact poorly correlated or uncorrelated with solar activity, and thus are of little value for predicting activity maxima. Also presented is a scheme in which the predictions of the individual proxies are combined via an appropriately weighted mean to produce a compound prediction. The scheme is then applied to the current cycle 22, and a maximum smoothed international sunspot number of 171 + or - 26 is estimated.

  10. The global significance of omitting soil erosion from soil organic carbon cycling schemes

    NASA Astrophysics Data System (ADS)

    Chappell, Adrian; Baldock, Jeffrey; Sanderman, Jonathan

    2016-02-01

    Soil organic carbon (SOC) cycling schemes used in land surface models (LSMs) typically account only for the effects of net primary production and heterotrophic respiration. To demonstrate the significance of omitting soil redistribution in SOC accounting, sequestration and emissions, we modified the SOC cycling scheme RothC (ref. ) to include soil erosion. Net SOC fluxes with and without soil erosion for Australian long-term trial sites were established and estimates made across Australia and other global regions based on a validated relation with catchment-scale soil erosion. Assuming that soil erosion is omitted from previous estimates of net C flux, we found that SOC erosion is incorrectly attributed to respiration. On this basis, the Australian National Greenhouse Gas inventory overestimated the net C flux from cropland by up to 40% and the potential (100 year) C sink is overestimated by up to 17%. We estimated global terrestrial SOC erosion to be 0.3-1.0 Pg C yr-1 indicating an uncertainty of -18 to -27% globally and +35 to -82% regionally relative to the long-term (2000-2010) terrestrial C flux of several LSMs. Including soil erosion in LSMs should reduce uncertainty in SOC flux estimates with implications for CO2 emissions, mitigation and adaptation strategies and interpretations of trends and variability in global ecosystems.

  11. Economic Analyiss of "Symbiotic" Light Water Reactor/Fast Burner Reactor Fuel Cycles Proposed as Part of the U.S. Advanced Fuel Cycle Initiative (AFCI)

    SciTech Connect

    Williams, Kent Alan; Shropshire, David E.

    2009-01-01

    A spreadsheet-based 'static equilibrium' economic analysis was performed for three nuclear fuel cycle scenarios, each designed for 100 GWe-years of electrical generation annually: (1) a 'once-through' fuel cycle based on 100% LWRs fueled by standard UO2 fuel assemblies with all used fuel destined for geologic repository emplacement, (2) a 'single-tier recycle' scenario involving multiple fast burner reactors (37% of generation) accepting actinides (Pu,Np,Am,Cm) from the reprocessing of used fuel from the uranium-fueled LWR fleet (63% of generation), and (3) a 'two-tier' 'thermal+fast' recycle scenario where co-extracted U,Pu from the reprocessing of used fuel from the uranium-fueled part of the LWR fleet (66% of generation) is recycled once as full-core LWR MOX fuel (8% of generation), with the LWR MOX used fuel being reprocessed and all actinide products from both UO2 and MOX used fuel reprocessing being introduced into the closed fast burner reactor (26% of generation) fuel cycle. The latter two 'closed' fuel cycles, which involve symbiotic use of both thermal and fast reactors, have the advantages of lower natural uranium requirements per kilowatt-hour generated and less geologic repository space per kilowatt-hour as compared to the 'once-through' cycle. The overall fuel cycle cost in terms of $ per megawatt-hr of generation, however, for the closed cycles is 15% (single tier) to 29% (two-tier) higher than for the once-through cycle, based on 'expected values' from an uncertainty analysis using triangular distributions for the unit costs for each required step of the fuel cycle. (The fuel cycle cost does not include the levelized reactor life cycle costs.) Since fuel cycle costs are a relatively small percentage (10 to 20%) of the overall busbar cost (LUEC or 'levelized unit electricity cost') of nuclear power generation, this fuel cycle cost increase should not have a highly deleterious effect on the competitiveness of nuclear power. If the reactor life cycle

  12. Influence of data collection schemes on the Life Cycle Assessment of a municipal wastewater treatment plant.

    PubMed

    Yoshida, Hiroko; Clavreul, Julie; Scheutz, Charlotte; Christensen, Thomas H

    2014-06-01

    A Life Cycle Assessment (LCA) of a municipal wastewater treatment plant (WWTP) was conducted to illustrate the effect of an emission inventory data collection scheme on the outcomes of an environmental impact assessment. Due to their burden in respect to data collection, LCAs often rely heavily on existing emission and operational data, which are gathered under either compulsory monitoring or reporting requirements under law. In this study, an LCA was conducted using three input data sources: Information compiled under compulsory disclosure requirements (the European Pollutant Release and Transfer Registry), compliance with national discharge limits, and a state-of-the-art emission data collection scheme conducted at the same WWTP. Parameter uncertainty for each collection scheme was assessed through Monte Carlo simulation. The comparison of the results confirmed that LCA results depend heavily on input data coverage. Due to the threshold on reporting value, the E-PRTR did not capture the impact for particulate matter emission, terrestrial acidification, or terrestrial eutrophication. While the current practice can capture more than 90% of non-carcinogenic human toxicity and marine eutrophication, an LCA based on the data collection scheme underestimates impact potential due to limitations of substance coverage. Besides differences between data collection schemes, the results showed that 3-13,500% of the impacts came from background systems, such as from the provisioning of fuel, electricity, and chemicals, which do not need to be disclosed currently under E-PRTR. The incidental release of pollutants was also assessed by employing a scenario-based approach, the results of which demonstrated that these non-routine emissions could increase overall WWTP greenhouse gas emissions by between 113 and 210%. Overall, current data collection schemes have the potential to provide standardized data collection and form the basis for a sound environmental impact assessment, but

  13. Closing nuclear fuel cycle with fast reactors: problems and prospects

    SciTech Connect

    Shadrin, A.; Dvoeglazov, K.; Ivanov, V.

    2013-07-01

    The closed nuclear fuel cycle (CNFC) with fast reactors (FR) is the most promising way of nuclear energetics development because it prevents spent nuclear fuel (SNF) accumulation and minimizes radwaste volume due to minor actinides (MA) transmutation. CNFC with FR requires the elaboration of safety, environmentally acceptable and economically effective methods of treatment of SNF with high burn-up and low cooling time. The up-to-date industrially implemented SNF reprocessing technologies based on hydrometallurgical methods are not suitable for the reprocessing of SNF with high burn-up and low cooling time. The alternative dry methods (such as electrorefining in molten salts or fluoride technologies) applicable for such SNF reprocessing have not found implementation at industrial scale. So the cost of SNF reprocessing by means of dry technologies can hardly be estimated. Another problem of dry technologies is the recovery of fissionable materials pure enough for dense fuel fabrication. A combination of technical solutions performed with hydrometallurgical and dry technologies (pyro-technology) is proposed and it appears to be a promising way for the elaboration of economically, ecologically and socially accepted technology of FR SNF management. This paper deals with discussion of main principle of dry and aqueous operations combination that probably would provide safety and economic efficiency of the FR SNF reprocessing. (authors)

  14. Analysis of supercritical CO{sub 2} cycle control strategies and dynamic response for Generation IV Reactors.

    SciTech Connect

    Moisseytsev, A.; Sienicki, J. J.

    2011-04-12

    efficiency of 49.3 %. The other approach involves reducing the minimum cycle pressure significantly below the critical pressure such that the temperature drop in the turbine is increased while the minimum cycle temperature is maintained above the critical temperature to prevent the formation of a liquid phase. The latter approach also involves the addition of a precooler and a third compressor before the main compressor to retain the benefits of compression near the critical point with the main compressor. For a minimum cycle pressure of 1 MPa, a cycle efficiency of 49.5% is achieved. Either approach opens up the door to applying the SCO{sub 2} cycle to the VHTR. In contrast, the SFR system typically has a core outlet-inlet temperature difference of about 150 C such that the standard recompression cycle is ideally suited for direct application to the SFR. The ANL Plant Dynamics Code has been modified for application to the VHTR and SFR when the reactor side dynamic behavior is calculated with another system level computer code such as SAS4A/SYSSYS-1 in the SFR case. The key modification involves modeling heat exchange in the RHX, accepting time dependent tabular input from the reactor code, and generating time dependent tabular input to the reactor code such that both the reactor and S-CO{sub 2} cycle sides can be calculated in a convergent iterative scheme. This approach retains the modeling benefits provided by the detailed reactor system level code and can be applied to any reactor system type incorporating a S-CO{sub 2} cycle. This approach was applied to the particular calculation of a scram scenario for a SFR in which the main and intermediate sodium pumps are not tripped and the generator is not disconnected from the electrical grid in order to enhance heat removal from the reactor system thereby enhancing the cooldown rate of the Na-to-CO{sub 2} RHX. The reactor side is calculated with SAS4A/SASSYS-1 while the S-CO{sub 2} cycle is calculated with the Plant Dynamics

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

    SciTech Connect

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

    2016-01-01

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

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

    SciTech Connect

    Chang Oh

    2004-07-01

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

  17. Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

    NASA Astrophysics Data System (ADS)

    Fic, Adam; Składzień, Jan; Gabriel, Michał

    2015-03-01

    Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle), which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle). The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

  18. Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

    PubMed

    Hill, R N; Nutt, W M; Laidler, J J

    2011-01-01

    The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. PMID:21399407

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

    NASA Astrophysics Data System (ADS)

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

    2014-02-01

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

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

    SciTech Connect

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

    2014-02-12

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

  1. Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

    SciTech Connect

    Benedict, R.W.; Goff, K.M.

    1993-01-01

    The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations.

  2. Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

    SciTech Connect

    Benedict, R.W.; Goff, K.M.

    1993-03-01

    The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations.

  3. Economic prospects of the Integral Fast Reactor (IFR) fuel cycle

    SciTech Connect

    Chang, Y.I.; Till, C.E.

    1991-01-01

    The IFR fuel cycle based on pyroprocessing involves only few operational steps and the batch-oriented process equipment systems are compact. This results in major cost reductions in all of three areas of reprocessing, fabrication, and waste treatment. This document discusses the economic aspects of this fuel cycle.

  4. Behavior of actinides in the Integral Fast Reactor fuel cycle

    SciTech Connect

    Courtney, J.C.; Lineberry, M.J.

    1994-06-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors` confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

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

    NASA Astrophysics Data System (ADS)

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

    2015-09-01

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

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

    SciTech Connect

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

    2015-09-30

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

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

    SciTech Connect

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

    2012-06-06

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

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

    NASA Astrophysics Data System (ADS)

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

    2012-06-01

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

  9. Effect of reduced enrichment on the fuel cycle for research reactors

    SciTech Connect

    Travelli, A.

    1982-01-01

    The new fuels developed by the RERTR Program and by other international programs for application in research reactors with reduced uranium enrichment (<20% EU) are discussed. It is shown that these fuels, combined with proper fuel-element design and fuel-management strategies, can provide at least the same core residence time as high-enrichment fuels in current use, and can frequently significantly extend it. The effect of enrichment reduction on other components of the research reactor fuel cycle, such as uranium and enrichment requirements, fuel fabrication, fuel shipment, and reprocessing are also briefly discussed with their economic implications. From a systematic comparison of HEU and LEU cores for the same reference research reactor, it is concluded that the new fuels have a potential for reducing the research reactor fuel cycle costs while reducing, at the same time, the uranium enrichment of the fuel.

  10. Neutronics and Depletion Methods for Parametric Studies of Fluoride Salt Cooled High Temperature Reactors with Slab Fuel Geometry and Multi-Batch Fuel Management Schemes

    SciTech Connect

    Cisneros, Anselmo T.; Ilas, Dan

    2013-01-01

    The Advanced High-Temperature Reactor (AHTR) is a 3400 MWth fluoride-salt-cooled high-temperature reactor (FHR) that uses TRISO particle fuel compacted into slabs rather than spherical or cylindrical fuel compacts. Simplified methods are required for parametric design studies such that analyzing the entire feasible design space for an AHTR is tractable. These simplifications include fuel homogenization techniques to increase the speed of neutron transport calculations in depletion analysis and equilibrium depletion analysis methods to analyze systems with multi-batch fuel management schemes. This paper presents three elements of significant novelty. First, the Reactivity-Equivalent Physical Transformation (RPT) methodology usually applied in systems with coated-particle fuel in cylindrical and spherical geometries has been extended to slab geometries. Secondly, based on this newly developed RPT method for slab geometries, a methodology that uses Monte Carlo depletion approaches was further developed to search for the maximum discharge burnup in a multi-batch system by iteratively estimating the beginning of equilibrium cycle (BOEC) composition and sampling different discharge burnups. This Iterative Equilibrium Depletion Search (IEDS) method fully defines an equilibrium fuel cycle (keff, power, flux, and composition evolutions) but is computationally demanding, although feasible on single-processor workstations. Finally, an analytical method, the Non-Linear Reactivity Model, was developed by expanding the linear reactivity model to include an arbitrary number of higher order terms so that single-batch depletion results could be extrapolated to estimate the maximum discharge burnup and BOEC keff in systems with multi-batch fuel management schemes. Results from this method were benchmarked against equilibrium depletion analysis results using the IEDS.

  11. Synfuels from fusion: using the tandem mirror reactor and a thermochemical cycle to produce hydrogen

    SciTech Connect

    Werner, R.W.

    1982-11-01

    This study is concerned with the following area: (1) the tandem mirror reactor and its physics; (2) energy balance; (3) the lithium oxide canister blanket system; (4) high-temperature blanket; (5) energy transport system-reactor to process; (6) thermochemical hydrogen processes; (7) interfacing the GA cycle; (8) matching power and temperature demands; (9) preliminary cost estimates; (10) synfuels beyond hydrogen; and (11) thermodynamics of the H/sub 2/SO/sub 4/-H/sub 2/O system. (MOW)

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

    SciTech Connect

    Hans D. Gougar; Frederik Reitsma; Wessel Joubert

    2009-05-01

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

  13. Gas Core Reactor with Magnetohydrodynamic Power System and Cascading Power Cycle

    SciTech Connect

    Smith, Blair M.; Anghaie, Samim

    2004-03-15

    The U.S. Department of Energy initiative Generation IV aim is to produce an entire nuclear energy production system with next-generation features for certification before 2030. A Generation IV-capable system must have superior sustainability, safety and reliability, and economic cost advantages in comparison with third generation light water reactors (LWRs). A gas core reactor (GCR) with magnetohydrodynamic (MHD) power converter and cascading power cycle forms the basis for a Generation IV concept that is expected to set the upper performance limits in sustainability and power conversion efficiency among all existing and proposed fission powered systems. A gaseous core reactor delivering thousands of megawatt fission power acts as the heat source for a high-temperature MHD power converter. A uranium tetrafluoride fuel mix, with {approx}95% mol fraction helium gas, provides a stable working fluid for the primary MHD Brayton cycle. The hot working fluid exiting a topping cycle MHD generator has sufficient heat to drive a conventional helium Brayton cycle with 35% thermal efficiency as well as a superheated steam Rankine cycle, with up to 40% efficiency, which recovers the waste heat from the intermediate Brayton cycle. A combined cycle efficiency of close to 70% can be achieved with only a modest MHD topping cycle efficiency. The high-temperature direct-energy conversion capability of an MHD dynamo combined with an already sophisticated steam-powered turbine industry knowledge base allows the cascading cycle design to achieve breakthrough first-law energy efficiencies previously unheard of in the nuclear power industry. Although simple in concept, the gas core reactor design has not achieved the state of technological maturity that established high-temperature gas-cooled reactors and high-temperature molten salt core reactors have pioneered. However, the GCR-MHD concept has considerable promise; for example, like molten salt reactors the fuel is continuously cycled

  14. Hybrid fusion-fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

    NASA Astrophysics Data System (ADS)

    Shmelev, A. N.; Kulikov, G. G.; Kurnaev, V. A.; Salahutdinov, G. H.; Kulikov, E. G.; Apse, V. A.

    2015-12-01

    Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the 231Pa-232U-233U-Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of 232U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

  15. Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

    SciTech Connect

    Shmelev, A. N. Kulikov, G. G. Kurnaev, V. A. Salahutdinov, G. H. Kulikov, E. G. Apse, V. A.

    2015-12-15

    Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the {sup 231}Pa–{sup 232}U–{sup 233}U–Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of {sup 232}U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

  16. A Supercritical CO{sub 2} Cycle- a Promising Power Conversion System for Generation IV Reactors

    SciTech Connect

    Hejzlar, Pavel; Dostal, Vaclav; Driscoll, Michael J.

    2006-07-01

    Advances in power conversion systems (PCS) for Generation IV power plants are of high importance because of their impact on plant specific capital cost reduction, which can be more significant than the cost savings achieved through the modifications of the nuclear island itself. One such PCS candidate, especially attractive for reactor outlet temperatures in the range of 550 to 650 deg C, is applicable to lead-alloy, sodium, or liquid salt-cooled reactors, as well as direct-cycle CO{sub 2} cooled reactors. The efficiencies achievable in this medium temperature range exceed those of conventional Brayton cycles and supercritical steam Rankine cycles and are comparable to those of conventional helium Brayton cycles at turbine inlet temperatures of 800 to 900 deg C. The S-CO{sub 2} recompression cycle under evaluation at MIT, is described with its advantages, drawbacks and R and D needs. The cycle is shown to excel in efficiency, simplicity and compactness which projects to cost savings, and in lower sensitivity of efficiency to core bypass flow, component pressure losses and flow maldistribution in recuperators. On the other hand, the cycle is highly recuperative and thus requires very compact heat exchangers, poses challenges to design of piping for large units, and its control and part load operation is more complicated. (authors)

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

    SciTech Connect

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

    2005-08-15

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

  18. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    SciTech Connect

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D.

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

  19. Gas Core Reactor-MHD Power System with Cascading Power Cycle

    SciTech Connect

    Smith, Blair M.; Anghaie, Samim; Knight, Travis W.

    2002-07-01

    The US Department of Energy initiative Gen-IV aim is to produce an entire nuclear energy production system with next generation features for certification before 2030. A Generation 4 capable system must have superior sustainability, safety and reliability, and economic cost advantages in comparison with third generation light water reactors. A gas core reactor (GCR) with magnetohydrodynamic (MHD) power converter and cascading power cycle forms the basis for a Generation IV concept that is expected to set the upper performance limits in sustainability and power conversion efficiency among all existing and proposed fission powered systems. A gaseous core reactor delivering 1000's MW fission power acts as the heat source for a high temperature magnetohydrodynamic power converter. A uranium tetrafluoride fuel mix, with {approx}95% mole fraction helium gas, provides a stable working fluid for the primary MHD-Brayton cycle. A helium Brayton cycle extracts waste heat from the MHD generator with about 20% energy efficiency, but the low temperature side is still hot enough ({approx}1600 K) to drive a second conventional helium Brayton cycle with about 35% efficiency. There is enough heat at the low temperature side of the He-Brayton cycle to generate steam, and so another heat recovery cycle can be added, this time a Rankine steam cycle with up to 40% efficiency. The proof of concept does not require a tremendously efficient (first law) MHD cycle, the high temperature direct energy conversion capability of an MHD dynamo, combined with already sophisticated steam powered turbine industry knowledge base allows the cascading cycle design to achieve break-through first law energy efficiencies previously unheard of in the nuclear power industry. Although simple in concept, the gas core reactor design has not achieved the state of technological maturity that, say, molten salt or high-temperature gas-cooled reactors have pioneered. However, even on paper the GCR-MHD concept holds

  20. Hybrid sulfur cycle operation for high-temperature gas-cooled reactors

    SciTech Connect

    Gorensek, Maximilian B

    2015-02-17

    A hybrid sulfur (HyS) cycle process for the production of hydrogen is provided. The process uses a proton exchange membrane (PEM) SO.sub.2-depolarized electrolyzer (SDE) for the low-temperature, electrochemical reaction step and a bayonet reactor for the high-temperature decomposition step The process can be operated at lower temperature and pressure ranges while still providing an overall energy efficient cycle process.

  1. A Gas-Cooled-Reactor Closed-Brayton-Cycle Demonstration with Nuclear Heating

    SciTech Connect

    Lipinski, Ronald J.; Wright, Steven A.; Dorsey, Daniel J.; Williamson, Joshua; Peters, Curtis D.; Brown, Nicholas; Jablonski, Jennifer

    2005-02-06

    A gas-cooled reactor may be coupled directly to turbomachinery to form a closed-Brayton-cycle (CBC) system in which the CBC working fluid serves as the reactor coolant. Such a system has the potential to be a very simple and robust space-reactor power system. Gas-cooled reactors have been built and operated in the past, but very few have been coupled directly to the turbomachinery in this fashion. In this paper we describe the option for testing such a system with a small reactor and turbomachinery at Sandia National Laboratories. Sandia currently operates the Annular Core Research Reactor (ACRR) at steady-state powers up to 4 MW and has an adjacent facility with heavy shielding in which another reactor recently operated. Sandia also has a closed-Brayton-Cycle test bed with a converted commercial turbomachinery unit that is rated for up to 30 kWe of power. It is proposed to construct a small experimental gas-cooled reactor core and attach this via ducting to the CBC turbomachinery for cooling and electricity production. Calculations suggest that such a unit could produce about 20 kWe, which would be a good power level for initial surface power units on the Moon or Mars. The intent of this experiment is to demonstrate the stable start-up and operation of such a system. Of particular interest is the effect of a negative temperature power coefficient as the initially cold Brayton gas passes through the core during startup or power changes. Sandia's dynamic model for such a system would be compared with the performance data. This paper describes the neutronics, heat transfer, and cycle dynamics of this proposed system. Safety and radiation issues are presented. The views expressed in this document are those of the author and do not necessarily reflect agreement by the government.

  2. A Gas-Cooled-Reactor Closed-Brayton-Cycle Demonstration with Nuclear Heating

    NASA Astrophysics Data System (ADS)

    Lipinski, Ronald J.; Wright, Steven A.; Dorsey, Daniel J.; Peters, Curtis D.; Brown, Nicholas; Williamson, Joshua; Jablonski, Jennifer

    2005-02-01

    A gas-cooled reactor may be coupled directly to turbomachinery to form a closed-Brayton-cycle (CBC) system in which the CBC working fluid serves as the reactor coolant. Such a system has the potential to be a very simple and robust space-reactor power system. Gas-cooled reactors have been built and operated in the past, but very few have been coupled directly to the turbomachinery in this fashion. In this paper we describe the option for testing such a system with a small reactor and turbomachinery at Sandia National Laboratories. Sandia currently operates the Annular Core Research Reactor (ACRR) at steady-state powers up to 4 MW and has an adjacent facility with heavy shielding in which another reactor recently operated. Sandia also has a closed-Brayton-Cycle test bed with a converted commercial turbomachinery unit that is rated for up to 30 kWe of power. It is proposed to construct a small experimental gas-cooled reactor core and attach this via ducting to the CBC turbomachinery for cooling and electricity production. Calculations suggest that such a unit could produce about 20 kWe, which would be a good power level for initial surface power units on the Moon or Mars. The intent of this experiment is to demonstrate the stable start-up and operation of such a system. Of particular interest is the effect of a negative temperature power coefficient as the initially cold Brayton gas passes through the core during startup or power changes. Sandia's dynamic model for such a system would be compared with the performance data. This paper describes the neutronics, heat transfer, and cycle dynamics of this proposed system. Safety and radiation issues are presented. The views expressed in this document are those of the author and do not necessarily reflect agreement by the government.

  3. A dynamic fuel cycle analysis for a heterogeneous thorium-DUPIC recycle in CANDU reactors

    SciTech Connect

    Jeong, C. J.; Park, C. J.; Choi, H.

    2006-07-01

    A heterogeneous thorium fuel recycle scenario in a Canada deuterium uranium (CANDU) reactor has been analyzed by the dynamic analysis method. The thorium recycling is performed through a dry process which has a strong proliferation resistance. In the fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides, and fission products of a multiple thorium recycling fuel cycle were estimated and compared to those of the once-through fuel cycle. The analysis results have shown that the heterogeneous thorium fuel cycle can be constructed through the dry process technology. It is also shown that the heterogeneous thorium fuel cycle can reduce the spent fuel inventory and save on the natural uranium resources when compared with the once-through cycle. (authors)

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

    SciTech Connect

    Middleton, Bobby; Pasch, James Jay; Kruizenga, Alan Michael; Walker, Matthew

    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

  5. 75 FR 36648 - Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technologies Subcommittee

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-06-28

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technologies Subcommittee AGENCY: Office of Nuclear Energy, DOE. ACTION: Notice of open meeting correction. On June 21, 2010, the Department of Energy published a notice...

  6. 75 FR 61139 - Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technology Subcommittee

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-10-04

    ...This notice announces an open meeting of the Reactor and Fuel Cycle Technology (RFCT) Subcommittee. The RFCT Subcommittee is a subcommittee of the Blue Ribbon Commission on America's Nuclear Future (the Commission). The establishment of subcommittees is authorized in the Commission's charter. The Commission was organized pursuant to the Federal Advisory Committee Act (Pub. L. 92-463, 86 Stat.......

  7. Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

    SciTech Connect

    Professor Neill Todreas

    2001-10-01

    A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team indicates

  8. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    SciTech Connect

    Shropshire, D.E.; Herring, J.S.

    2004-10-03

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim

  9. Reactor applications of the Compact Fusion Advanced Rankine (CFAR) cycle for a D-T tokamak fusion reactor

    NASA Astrophysics Data System (ADS)

    Hoffman, H. A.; Logan, B. G.; Campbell, R. B.

    1988-03-01

    A preliminary design of a D-T fusion reactor blanket and MHD power conversion system is made based on the CFAR concept, and it was found that performance and costs for the reference cycle are very attractive. While much remains to be done, the potential advantage of liquid metal Rankine cycles for fusion applications are much clearer now. These include low pressures and mass flow rates, a nearly isothermal module shell which minimizes problems of thermal distortion and stresses, and an insensitivity to pressure losses in the blanket so that the two-phase MHD pressure drops in the boiling part of the blanket and the ordinary vapor pressure drops in the pebble-bed superheating zones are acceptable (the direct result of pumping a liquid rather than having to compress a gas). There are no moving parts in the high-temperature MHD power generators, no steam bottoming plant is required, only small vapor precoolers and condensers are needed because of the high heat rejection temperatures, and only a relatively small natural-draft heat exchanger is required to reject the heat to the atmosphere. The net result is a very compact fusion reactor and power conversion system which fit entirely inside an 18 meter radius reactor vault. Although a cost analysis has not yet been performed, preliminary cost estimates indicate low capital costs and a very attractive cost of electricity.

  10. Synfuels from fusion: producing hydrogen with the tandem mirror reactor and thermochemical cycles

    SciTech Connect

    Ribe, F.L.; Werner, R.W.

    1981-01-21

    This report examines, for technical merit, the combination of a fusion reactor driver and a thermochemical plant as a means for producing synthetic fuel in the basic form of hydrogen. We studied: (1) one reactor type - the Tandem Mirror Reactor - wishing to use to advantage its simple central cell geometry and its direct electrical output; (2) two reactor blanket module types - a liquid metal cauldron design and a flowing Li/sub 2/O solid microsphere pellet design so as to compare the technology, the thermal-hydraulics, neutronics and tritium control in a high-temperature operating mode (approx. 1200 K); (3) three thermochemical cycles - processes in which water is used as a feedstock along with a high-temperature heat source to produce H/sub 2/ and O/sub 2/.

  11. Driver options and burn-cycle selection based on power-reactor considerations

    SciTech Connect

    Ehst, D.A.

    1983-04-01

    Reactor implications for noninductive current drive are presented based on a number of studies. First, the lower hybrid driver for the STARFIRE reactor is discussed and the disadvantages of this driver are reviewed. Next, the results of an extensive search for a better current driver are presented. A large number of alternatives were compared in a common context, the DEMO reactor, in order to examine their suitability on a standard basis. Finally, the methodology of a study, currently in progress, is described. The goals of this last study are to compare tokamak reactor designs optimized for operation under different burn cycles, in order to assess the actual benefits and costs of pulsed versus steady-state operation.

  12. Evaluation of Indirect Combined Cycle in Very High Temperature Gas--Cooled Reactor

    SciTech Connect

    Chang Oh; Robert Barner; Cliff Davis; Steven Sherman; Paul Pickard

    2006-10-01

    The U.S. Department of Energy and Idaho National Laboratory are developing a very high temperature reactor to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is twofold: (a) efficient, low-cost energy generation and (b) hydrogen production. Although a next-generation plant could be developed as a single-purpose facility, early designs are expected to be dual purpose, as assumed here. A dual-purpose design with a combined cycle of a Brayton top cycle and a bottom Rankine cycle was investigated. An intermediate heat transport loop for transporting heat to a hydrogen production plant was used. Helium, CO2, and a helium-nitrogen mixture were studied to determine the best working fluid in terms of the cycle efficiency. The relative component sizes were estimated for the different working fluids to provide an indication of the relative capital costs. The relative size of the turbomachinery was measured by comparing the power input/output of the component. For heat exchangers the volume was computed and compared. Parametric studies away from the baseline values of the cycle were performed to determine the effects of varying conditions in the cycle. This gives some insight into the sensitivity of the cycle to various operating conditions as well as trade-offs between efficiency and component size. Parametric studies were carried out on reactor outlet temperature, mass flow, pressure, and turbine cooling.

  13. Design of a Solar Reactor to Split CO2 Via Isothermal Redox Cycling of Ceria

    SciTech Connect

    Bader, R; Chandran, RB; Venstrom, LJ; Sedler, SJ; Krenzke, PT; De Smith, RM; Banerjee, A; Chase, TR; Davidson, JH; Lipinski, W

    2014-12-23

    The design procedure for a 3 kWth prototype solar thermochemical reactor to implement isothermal redox cycling of ceria for CO2 splitting is presented. The reactor uses beds of mm-sized porous ceria particles contained in the annulus of concentric alumina tube assemblies that line the cylindrical wall of a solar cavity receiver. The porous particle beds provide high surface area for the heterogeneous reactions, rapid heat and mass transfer, and low pressure drop. Redox cycling is accomplished by alternating flows of inert sweep gas and CO2 through the bed. The gas flow rates and cycle step durations are selected by scaling the results from small-scale experiments. Thermal and thermo-mechanical models of the reactor and reactive element tubes are developed to predict the steady-state temperature and stress distributions for nominal operating conditions. The simulation results indicate that the target temperature of 1773K will be reached in the prototype reactor and that the Mohr-Coulomb static factor of safety is above two everywhere in the tubes, indicating that thermo-mechanical stresses in the tubes remain acceptably low.

  14. Comparative study of pulsed and steady-state tokamak reactor burn cycles

    SciTech Connect

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K.; Hassanein, A.M.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1984-05-01

    Four distinct operating modes have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue in pulsed poloidal field coils; out-of-plant fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics and engineering which will help achieve these goals for different burn cycles.

  15. Comparison of accelerator-based with reactor-based waste transmutation schemes

    SciTech Connect

    Sailor, W.C.; Beard, C.A.; Venneri, F.; Davidson, J.W.

    1993-12-01

    Accelerator-based transmutation of waste (ATW) systems for the destruction of commercial LWR spent fuel are compared with systems based on thermal reactors accomplish the same objectives. When the same technology is assumed for the actinide-burning aspect of the two systems, it is seen that the size of the accelerator is determined only by the choice of how many of the long-lived fission products to burn. if none are transmuted, then the accelerator is not necessary. This result is independent of the choice of fluid carrier, and whether the actinides are destroyed in an ATW system or in a separate reactor.

  16. Phase Cycling Schemes for finite-pulse-RFDR MAS Solid State NMR Experiments

    PubMed Central

    Zhang, Rongchun; Nishiyama, Yusuke; Sun, Pingchuan; Ramamoorthy, Ayyalusamy

    2015-01-01

    The finite-pulse radio frequency driven dipolar recoupling (fp-RFDR) pulse sequence is used in 2D homonuclear chemical shift correlation experiments under magic angle spinning (MAS). A recent study demonstrated the advantages of using a short phase cycle, XY4, and its super-cycle, XY414, for the fp-RFDR pulse sequence employed in 2D 1H/1H single-quantum/single-quantum correlation experiments under ultrafast MAS conditions. In this study, we report a comprehensive analysis on the dipolar recoupling efficiencies of XY4, XY412, XY413, XY414, and XY814 phase cycles under different spinning speeds ranging from 10 to 100 kHz. The theoretical calculations reveal the presence of second-order terms (T10T2,±2, T1,±1T2,±1, etc.) in the recoupled homonuclear dipolar coupling Hamiltonian only when the basic XY4 phase cycle is utilized, making it advantageous for proton-proton magnetization transfer under ultrafast MAS conditions. It is also found that the recoupling efficiency of fp-RFDR is quite dependent on the duty factor (τ180/τR) as well as on the strength of homonuclear dipolar couplings. The rate of longitudinal magnetization transfer increases linearly with the duty factor of fp-RFDR for all the XY-based phase cycles investigated in this study. Examination of the performances of different phase cycles against chemical shift offset and RF field in homogeneity effects revealed that XY414 is the most tolerant phase cycle, while the shortest phase cycle XY4 suppressed the RF field inhomogeneity effects most efficiently under slow spinning speeds. Our results suggest that the difference in the fp-RFDR recoupling efficiencies decreases with the increasing MAS speed, while ultrafast (>60 kHz) spinning speed is advantageous as it recouples a large amount of homonuclear dipolar couplings and therefore enable fast magnetization exchange. The effects of higher-order terms and cross terms between various interactions in the effective Hamiltonian of fp-RFDR are also analyzed

  17. Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

    SciTech Connect

    Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A. Ignatiev, V. V.; Subbotin, S. A. Tsibulskiy, V. F.

    2015-12-15

    The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.

  18. Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

    NASA Astrophysics Data System (ADS)

    Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A.; Ignatiev, V. V.; Subbotin, S. A.; Tsibulskiy, V. F.

    2015-12-01

    The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.

  19. Apparatus and process to eliminate diffusional limitations in a membrane biological reactor by pressure cycling

    DOEpatents

    Efthymiou, George S.; Shuler, Michael L.

    1989-08-29

    An improved multilayer continuous biological membrane reactor and a process to eliminate diffusional limitations in membrane reactors in achieved by causing a convective flux of nutrient to move into and out of an immobilized biocatalyst cell layer. In a pressure cycled mode, by increasing and decreasing the pressure in the respective layers, the differential pressure between the gaseous layer and the nutrient layer is alternately changed from positive to negative. The intermittent change in pressure differential accelerates the transfer of nutrient from the nutrient layers to the biocatalyst cell layer, the transfer of product from the cell layer to the nutrient layer and the transfer of byproduct gas from the cell layer to the gaseous layer. Such intermittent cycling substantially eliminates mass transfer gradients in diffusion inhibited systems and greatly increases product yield and throughput in both inhibited and noninhibited systems.

  20. Impacts of Heterogeneous Recycle in Fast Reactors on Overall Fuel Cycle

    SciTech Connect

    Temitope A. Taiwo; Samuel E. Bays; Abdullatif M. Yacout; Edward M. Hoffman; Michael Todosow; Taek K. Kim; Massimo Salvatores

    2011-03-01

    A study in the United States has evaluated the attributes of the heterogeneous recycle approach for plutonium and minor actinide transmutation in fast reactor fuel cycles, with comparison to the homogeneous recycle approach, where pertinent. The work investigated the characteristics, advantages, and disadvantages of the approach in the overall fuel cycle, including reactor transmutation, systems and safety impacts, fuel separation and fabrication issues, and proliferation risk and transportation impacts. For this evaluation, data from previous and ongoing national studies on heterogeneous recycle were reviewed and synthesized. Where useful, information from international sources was included in the findings. The intent of the work was to provide a comprehensive assessment of the heterogeneous recycle approach at the current time.

  1. The benefits of a fast reactor closed fuel cycle in the UK

    SciTech Connect

    Gregg, R.; Hesketh, K.

    2013-07-01

    The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size, so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since the

  2. POWER CYCLE AND STRESS ANALYSES FOR HIGH TEMPERATURE GAS-COOLED REACTOR

    SciTech Connect

    Oh, Chang H; Davis, Cliff; Hawkes, Brian D; Sherman, Steven R

    2007-05-01

    The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. Many aspects of the NGNP must be researched and developed in order to make recommendations on the final design of the plant. Parameters such as working conditions, cycle components, working fluids, and power conversion unit configurations must be understood. Three configurations of the power conversion unit were demonstrated in this study. A three-shaft design with three turbines and four compressors, a combined cycle with a Brayton top cycle and a Rankine bottoming cycle, and a reheated cycle with three stages of reheat were investigated. An intermediate heat transport loop for transporting process heat to a High Temperature Steam Electrolysis (HTSE) hydrogen production plant was used. Helium, CO2, and a 80% nitrogen, 20% helium mixture (by weight) were studied to determine the best working fluid in terms cycle efficiency and development cost. In each of these configurations the relative component size were estimated for the different working fluids. The relative size of the turbomachinery was measured by comparing the power input/output of the component. For heat exchangers the volume was computed and compared. Parametric studies away from the baseline values of the three-shaft and combined cycles were performed to determine the effect of varying conditions in the cycle. This gives some insight into the sensitivity of these cycles to

  3. Boiling Water Reactor Fuel Cycle Optimization for Prevention of Channel-Blade Interference

    SciTech Connect

    Kropaczek, David J.; Karve, Atul A.; Oyarzun, Christian C.; Asgari, Mehdi

    2006-07-01

    A formal optimization method for eliminating the potential of Boiling Water Reactor channel-blade interference is presented within the context of fuel cycle design. The method is based on the use of threshold constraints on blade force as penalty terms within an objective function that are employed as part of a search algorithm. Results demonstrate the effectiveness of the constraint formulation in eliminating channel-blade interference as part of the design of the core loading and operational strategy. (authors)

  4. Generation of attosecond x-ray pulses with a multi-cycle two-color ESASE scheme

    SciTech Connect

    Ding, Y.; Huang, Z.; Ratner, D.; Bucksbaum, P.; Merdji, H.; /Saclay /SLAC

    2009-03-04

    Generation of attosecond x-ray pulses is attracting much attention within the x-ray free-electron laser (FEL) user community. Several schemes using extremely short laser pulses to manipulate the electron bunches have been proposed. In this paper, we extend the attosecond two-color ESASE scheme proposed by Zholents et al. to the long optical cycle regime using a second detuned laser and a tapered undulator. Both lasers can be about ten-optical-cycles long, with the second laser frequency detuned from the first to optimize the contrast between the central and side current spikes. A tapered undulator mitigates the degradation effect of the longitudinal space charge (LSC) force in the undulator and suppresses the FEL gain of all side current peaks. Simulations using the LCLS parameters show a single attosecond x-ray spike of {approx} 110 attoseconds can be produced. The second laser can also be detuned to coherently control the number of the side x-ray spikes and the length of the radiation pulse.

  5. Supercritical Carbon Dioxide Brayton Power Conversion Cycle Design for Optimized Battery-Type Integral Reactor System

    SciTech Connect

    Kim, Won J.; Kim, Tae W.; Sohn, Myoung S.; Suh, Kune Y.

    2006-07-01

    Supercritical carbon dioxide (SCO{sub 2}) promises a high power conversion efficiency of the recompression Brayton cycle due to its excellent compressibility reducing the compression work at the bottom of the cycle and to a higher density than helium or steam decreasing the component size. Therefore, the high SCO{sub 2} Brayton cycle efficiency as high as 45 % furnishes small sized nuclear reactors with economical benefits on the plant construction and maintenance. A 23 MWth BORIS (Battery Optimized Reactor Integral System) is being developed as a multipurpose reactor. BORIS, an integral-type optimized fast reactor with an ultra long life core, is coupled to the SCO{sub 2} Brayton cycle needing less room relative to the Rankine steam cycle because of its smaller components. The SCO{sub 2} Brayton cycle of BORIS consists of a 16 MW turbine, a 32 MW high temperature recuperator, a 14 MW low temperature recuperator, an 11 MW pre-cooler and 2 and 2.8 MW compressors. Entering six heat exchangers between primary and secondary system at 19.9 MPa and 663 K, the SCO{sub 2} leaves the heat exchangers at 19.9 MPa and 823 K. The promising secondary system efficiency of 45 % was calculated by a theoretical method in which the main parameters include pressure, temperature, heater power, the turbine's, recuperators' and compressors' efficiencies, and the flow split ratio of SCO{sub 2} going out from the low temperature recuperator. Test loop SOLOS (Shell-and-tube Overall Layout Optimization Study) is utilized to develop advanced techniques needed to adopt the shell-and-tube type heat exchanger in the secondary loop of BORIS by studying the SCO{sub 2} behavior from both thermal and hydrodynamic points of view. Concurrently, a computational fluid dynamics (CFD) code analysis is being conducted to develop an optimal analytical method of the SCO{sub 2} turbine efficiency having the parameters of flow characteristics of SCO{sub 2} passing through buckets of the turbine. These

  6. High Temperature Fusion Reactor Cooling Using Brayton Cycle Based Partial Energy Conversion

    NASA Technical Reports Server (NTRS)

    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.

  7. Impact of thorium based molten salt reactor on the closure of the nuclear fuel cycle

    NASA Astrophysics Data System (ADS)

    Jaradat, Safwan Qasim Mohammad

    Molten salt reactor (MSR) is one of six reactors selected by the Generation IV International Forum (GIF). The liquid fluoride thorium reactor (LFTR) is a MSR concept based on thorium fuel cycle. LFTR uses liquid fluoride salts as a nuclear fuel. It uses 232Th and 233U as the fertile and fissile materials, respectively. Fluoride salt of these nuclides is dissolved in a mixed carrier salt of lithium and beryllium (FLiBe). The objective of this research was to complete feasibility studies of a small commercial thermal LFTR. The focus was on neutronic calculations in order to prescribe core design parameter such as core size, fuel block pitch (p), fuel channel radius, fuel path, reflector thickness, fuel salt composition, and power. In order to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core reactor. The MCNP code was used to study the reactor physics characteristics for the FUJI-U3 reactor. The results were then compared with the results obtained from the original FUJI-U3 using the reactor physics code SRAC95 and the burnup analysis code ORIPHY2. The results were comparable with each other. Based on the results, MCNP was found to be a reliable code to model a small thermal LFTR and study all the related reactor physics characteristics. The results of this study were promising and successful in demonstrating a prefatory small commercial LFTR design. The outcome of using a small core reactor with a diameter/height of 280/260 cm that would operate for more than five years at a power level of 150 MWth was studied. The fuel system 7LiF - BeF2 - ThF4 - UF4 with a (233U/ 232Th) = 2.01 % was the candidate fuel for this reactor core.

  8. A master-follower type distributed scheme for reactor inlet temperature control

    SciTech Connect

    Garcia, H.E.; Dean, E.M.; Vilim, R.B.

    1995-06-01

    This paper describes the implementation of a computer-based controller for regulating reactor inlet temperature in a pool-type power plant. The elements of the control system are organized in a master-follower hierarchical architecture that takes advantage of existing in-plant hardware and software to minimize the need for plant modifications. Low level control algorithms are executed on existing local digital controllers (followers) with the high level algorithms executed on a new plant supervisory computer (master). A distributed computing strategy provides integration of the existing and additional computer platforms. The control system operates by having the master controller first estimate the secondary sodium flow needed to achieve a given reactor inlet temperature. The estimated flow is then used as a setpoint by the follower controller to regulate sodium flow using a motor-generator pump set. The control system has been implemented in a Hardware-In-the-Loop (FM) setup and qualified for operation in the Experimental Breader reactor 11 of Argonne National Laboratory. Some HIL results are provided.

  9. Optimization of the Mode of the Uranium-233 Accumulation for Application in Thorium Self-Sufficient Fuel Cycle of Candu Power Reactor

    SciTech Connect

    Bergelson, Boris; Gerasimov, Alexander; Tikhomirov, Georgy

    2006-07-01

    Results of calculation studies of the first stage of self-sufficient thorium cycle for CANDU reactor are presented in the paper. The first stage is preliminary accumulation of {sup 233}U in the CANDU reactor itself. Parameters of active core and scheme of fuel reloading were accepted the same as those for CANDU reactor. It was assumed for calculations, that enriched {sup 235}U or plutonium was used as additional fissile material to provide neutrons for {sup 233}U production. Parameters of 10 different variants of the elementary cell of active core were calculated for the lattice pitch, geometry of fuel channels, and fuel assembly of the CANDU reactor. The results presented in the paper allow to determine the time of accumulation of the required amount of {sup 233}U and corresponding number of targets going into processing for {sup 233}U extraction. Optimum ratio of the accumulation time to number of processed targets can be determined using the cost of electric power produced by the reactor and cost of targets along with their processing. (authors)

  10. Development and experimental validation of a calculation scheme for nuclear heating evaluation in the core of the OSIRIS material testing reactor

    SciTech Connect

    Malouch, F.

    2011-07-01

    The control of the temperature in material samples irradiated in a material testing reactor requires the knowledge of the nuclear heating caused by the energy deposition by neutrons and photons interacting in the irradiation device structures. Thus, a neutron-photonic three-dimensional calculation scheme has been developed to evaluate the nuclear heating in experimental devices irradiated in the core of the OSIRIS MTR reactor (CEA/Saclay Center). The aim is to obtain a predictive tool for the nuclear heating estimation in irradiation devices. This calculation scheme is mainly based on the TRIPOLI-4 three-dimensional continuous-energy Monte Carlo transport code, developed by CEA (Saclay Center). An experimental validation has been carried out on the basis of nuclear heating measurements performed in the OSIRIS core. After an overview of the experimental devices irradiated in the OSIRIS reactor, we present the calculation scheme and the first results of the experimental validation. (authors)

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

    NASA Astrophysics Data System (ADS)

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

    2016-03-01

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

  12. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 3-Surry Unit 1 Cycle 2

    SciTech Connect

    Bowman, S.M.

    1995-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using selected critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations in this report is based on the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and to provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of two reactor critical configurations for Surry Unit 1 Cycle 2. This unit and cycle were chosen for a previous analysis using a different methodology because detailed isotopics from multidimensional reactor calculations were available from the Virginia Power Company. These data permitted a direct comparison of criticality calculations using the utility-calculated isotopics with those using the isotopics generated by the SCALE-4

  13. Optimization strategies for sustainable fuel cycle of the BR2 Reactor

    SciTech Connect

    Kalcheva, S.; Van Den Branden, G.; Koonen, E.

    2013-07-01

    The objective of the present study is to achieve a sustainable fuel cycle in a long term of reactor operation applying advanced in-core loading strategies. The optimization criteria concern mainly enhancement of nuclear safety by means of reactivity margins and minimization of the operational fuel cycle cost at a given (constant) power level and same or longer cycle length. An important goal is also to maintain the same or to improve the experimental performances. Current developments are focused on optimization of control rods localization; optimization of fresh and burnt fuel assemblies in-core distribution; optimization of azimuth and axial fuel burn up strategies, including fuel assembly rotating and flipping upside down. (authors)

  14. A Novel Fuel/Reactor Cycle to Implement the 300 Years Nuclear Waste Policy Approach - 12377

    SciTech Connect

    Carelli, M.D.; Franceschini, F.; Lahoda, E.J.; Petrovic, B.

    2012-07-01

    A thorium-based fuel cycle system can effectively burn the currently accumulated commercial used nuclear fuel and move to a sustainable equilibrium where the actinide levels in the high level waste are low enough to yield a radiotoxicity after 300 years lower than that of the equivalent uranium ore. The second step of the Westinghouse approach to solving the waste 'problem' has been completed. The thorium fuel cycle has indeed the potential of burning the legacy TRU and achieve the waste objective proposed. Initial evaluations have been started for the third step, development and selection of appropriate reactors. Indications are that the probability of show-stoppers is rather remote. It is, therefore, believed that development of the thorium cycle and associated technologies will provide a permanent solution to the waste management. Westinghouse is open to the widest collaboration to make this a reality. (authors)

  15. Crosscutting Requirements in the International Project on Innovative Reactors and Fuel Cycles (INPRO)

    SciTech Connect

    Steur, Ronald; Lyubenov Yaven, Yanko; Gueorguiev, Boris; Mahadeva, Rao; Shen, Wenquan

    2002-07-01

    There are two categories of requirements: (i) user requirements that need to be met by the designers and manufacturers of innovative reactors and fuel cycles, and (ii) a wide spectrum of requirements that need to be met by countries, willing to successfully deploy innovative nuclear reactors for energy production. This part of the International Project on Innovative Reactors and Fuel Cycles will mainly deal with the second category of requirements. Both categories of requirements will vary depending on the institutional development, infrastructure availability and social attitude in any given country. Out of the need for sustainable development requirements will also more specific in the future. Over a 50-year time frame both categories of requirements will evolve with social and economic development as nuclear technology develops further. For example, the deployment of innovative reactors in countries with marginal or non-existing nuclear infrastructures would be possible only if the reactors are built, owned and operated by an international nuclear utility or if they are inherently safe and can be delivered as a 'black box - nuclear battery'. A number of issues will need to be addressed and conditions and requirements developed if this is going to become a reality. One general requirement for wider utilization of innovative nuclear power will be the public and environmental considerations, which will play a role in the decision making processes. Five main clusters of topics will be handled: - Infra-structural aspects, typology and consequences for nuclear development. - Industrial requirements for the different innovative concepts. - Institutional developments and requirements for future deployment of nuclear energy. (National as well as international) - Socio-political aspects, a.o. public acceptance and role of governments. - Sustainability: requirements following the need for sustainability Analysis will be made of the evolution of national and international

  16. MESMO 2: a mechanistic marine silica cycle and coupling to a simple terrestrial scheme

    NASA Astrophysics Data System (ADS)

    Matsumoto, K.; Tokos, K.; Huston, A.; Joy-Warren, H.

    2013-04-01

    Here we describe the second version of Minnesota Earth System Model for Ocean biogeochemistry (MESMO 2), an earth system model of intermediate complexity, which consists of a dynamical ocean, dynamic-thermodynamic sea ice, and energy moisture balanced atmosphere. The new version has more realistic land ice masks and is driven by seasonal winds. A major aim in version 2 is representing the marine silica cycle mechanistically in order to investigate climate-carbon feedbacks involving diatoms, a critically important class of phytoplankton in terms of carbon export production. This is achieved in part by including iron, on which phytoplankton uptake of silicic acid depends. Also, MESMO 2 is coupled to an existing terrestrial model, which allows for the exchange of carbon, water and energy between land and the atmosphere. The coupled model, called MESMO 2E, is appropriate for more complete earth system simulations. The new version was calibrated, with the goal of preserving reasonable interior ocean ventilation and various biological production rates in the ocean and land, while simulating key features of the marine silica cycle.

  17. MESMO 2: a mechanistic marine silica cycle and coupling to a simple terrestrial scheme

    NASA Astrophysics Data System (ADS)

    Matsumoto, K.; Tokos, K. S.; Huston, A.; Joy-Warren, H.

    2012-09-01

    Here we describe the second version of Minnesota Earth System Model for Ocean biogeochemistry (MESMO 2), an earth system model of intermediate complexity, which consists of a dynamical ocean, dynamic-thermodynamic sea ice, and energy moisture balanced atmosphere. The new version has more realistic land ice masks and is driven by seasonal winds. A major aim in version 2 is representing the marine silica cycle mechanistically in order to investigate climate-carbon feedbacks involving diatoms, a critically important class of phytoplankton in terms of carbon export production. This is achieved in part by including iron, on which phytoplankton uptake of silicic acid depends. Also, MESMO 2 is coupled to an existing terrestrial model, which allows for the exchange of carbon, water, and energy between land and the atmosphere. The coupled model, called MESMO 2E, is appropriate for more complete earth system simulations. The new version was calibrated with the goal of preserving reasonable interior ocean ventilation and various biological production rates in the ocean and land, while simulating key features of the marine silica cycle.

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

    SciTech Connect

    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.

  19. Scale-4 analysis of pressurized water reactor critical configurations: Volume 5, North Anna Unit 1 Cycle 5

    SciTech Connect

    Bowman, S.M.; Suto, T. |

    1996-10-01

    ANSI/ANS 8.1 requires that calculational methods for away-from- reactor (AFR) criticality safety analyses be validated against experiment. This report summarizes part of the ongoing effort to benchmark AFR criticality analysis methods using selected critical configurations from commercial PWRs. Codes and data in the SCALE-4 code system were used. This volume documents the SCALE system analysis of one reactor critical configuration for North Anna Unit 1 Cycle 5. The KENO V.a criticality calculations for the North Anna 1 Cycle 5 beginning-of-cycle model yielded a value for k{sub eff} of 1. 0040{+-}0.0005.

  20. Proliferation resistance of the fuel cycle for the Integral Fast Reactor

    SciTech Connect

    Burris, L.

    1993-09-01

    Argonne National Laboratory has developed an electrorefining pyrochemical process for recovery and recycle of metal fuel discharged from the Integral Fast Reactor (FR). This inherently low decontamination process has an overall decontamination factor of only about 100 for the plutonium metal product. As a result, all of the fuel cycle operations must be conducted in heavily shielded cells containing a high-purity argon atmosphere. The FR fuel cycle possesses high resistance to clandestine diversion or overt, state- supported removal of plutonium for nuclear weapons production because of two main factors: the highly radioactive product, which is also contaminated with heat- and neutron-producing isotopes of plutonium and other actinide elements, and the difficulty of removing material from the FR facility through the limited number of cell transfer locks without detection.

  1. Effect of changes in DOE pricing policies for enrichment and reprocessing on research reactor fuel cycle costs

    SciTech Connect

    Matos, J.E.; Freese, K.E.

    1986-11-03

    Fuel cycle costs with HEU and LEU fuels for the IAEA generic 10 MW reactor are updated to reflect the change in DOE pricing policy for enrichment services as of October 1985 and the published charges for LEU reprocessing services as of February 1986. The net effects are essentially no change in HEU fuel cycle costs and a reduction of about 8 to 10% in the fuel cycle costs for LEU silicide fuel.

  2. Advanced Fuel Cycles for Fusion Reactors: Passive Safety and Zero-Waste Options

    NASA Astrophysics Data System (ADS)

    Zucchetti, Massimo; Sugiyama, Linda E.

    2006-05-01

    Nuclear fusion is seen as a much ''cleaner'' energy source than fission. Most of the studies and experiments on nuclear fusion are currently devoted to the Deuterium-Tritium (DT) fuel cycle, since it is the easiest way to reach ignition. The recent stress on safety by the world's community has stimulated the research on other fuel cycles than the DT one, based on 'advanced' reactions, such as the Deuterium-Helium-3 (DHe) one. These reactions pose problems, such as the availability of 3He and the attainment of the higher plasma parameters that are required for burning. However, they have many advantages, like for instance the very low neutron activation, while it is unnecessary to breed and fuel tritium. The extrapolation of Ignitor technologies towards a larger and more powerful experiment using advanced fuel cycles (Candor) has been studied. Results show that Candor does reach the passive safety and zero-waste option. A fusion power reactor based on the DHe cycle could be the ultimate response to the environmental requirements for future nuclear power plants.

  3. Plant System Design of Supercritical CO{sub 2} Direct Cycle Gas Turbine Fast Reactor

    SciTech Connect

    Katsuhiro, Tozawa; Nobumasa, Tsuji; Yasushi, Muto; Yasuyoshi, Kato

    2006-07-01

    The conceptual plant design and preliminary safety analysis of SCDFR, Supercritical CO{sub 2} Direct Cycle Gas Turbine Fast Reactor, were performed. Plant thermal power is 600 MW. Core outlet/inlet pressure and temperature are 12.5/12.8 MPa and 527/388 deg C respectively. The core height and equivalent diameter are about 1.2 m and about 3.146 m respectively. The core can be burning for 10 years without refueling by adding 6.5% content of {sup 237}Np into the fuel as a burnable poison. Reactor pressure vessel height and inner diameter are about 19.3 m and about 6.55 m respectively. Steel containment vessel contains the reactor system and the gas turbine system. Preliminary analysis of core temperature behavior during the depressurization accident in SCDFR was performed. In the result of the analysis, core temperature is limited under 900 deg C, assumed limit temperature of the fuel clad, at the condition of minimum gas circulation flow rate of 2.0 m{sup 3}/s. On the other hand, gas circulator designed flow rate of the auxiliary core cooling system is over 11.6 m{sup 3}/s. These show that the integrity of the fuel clad during depressurization accident is maintained. We conclude that the plant concept of SCDFR is developed and the plant safety under depressurization accident conditions is confirmed by preliminary analysis. (authors)

  4. Treatment of sewage sludge in a thermophilic membrane reactor (TMR) with alternate aeration cycles.

    PubMed

    Collivignarelli, Maria Cristina; Castagnola, Federico; Sordi, Marco; Bertanza, Giorgio

    2015-10-01

    The management of sewage sludge is becoming a more and more important issue, both at national and international level, in particular due to the uncertain recovery/disposal future options. Therefore, it is clear that the development of new technologies that can mitigate the problem at the source by reducing sludge production is necessary, such as the European Directive 2008/98/EC prescribes. This work shows the results obtained with a thermophilic membrane reactor, for processing a biological sludge derived from a wastewater treatment plant (WWTP) that treats urban and industrial wastewater. Sewage sludge was treated in a thermophilic membrane reactor (TMR), at pilot-scale (1 m(3) volume), with alternate aeration cycles. The experimentation was divided into two phases: a "startup phase" during which, starting with a psychrophilic/mesophilic biomass, thermophilic conditions were progressively reached, while feeding a highly biodegradable substrate; the obtained thermophilic biomass was then used, in the "regime phase", to digest biological sludge which was fed to the plant. Good removal yields were observed: 64% and 57% for volatile solids (VS) and total COD (CODtot), respectively, with an average hydraulic retention time (HRT) equal to 20 d, an organic loading rate (OLR) of about 1.4-1.8 kg COD m(-3) d(-1) and aeration/non aeration cycles alternated every 4 h. PMID:26233586

  5. Measurement control design and performance assessment in the Integral Fast Reactor fuel cycle

    SciTech Connect

    Orechwa, Y.; Bucher, R.G.

    1994-08-01

    The Integral Fast Reactor (IFR)--consisting of a metal fueled and liquid metal cooled reactor together with an attendant fuel cycle facility (FCF)--is currently undergoing a phased demonstration of the closed fuel cycle at Argonne National Laboratory. The recycle technology is pyrometalurgical based with incomplete fission product separation and all transuranics following plutonium for recycle. The equipment operates in batch mode at 500 to 1,300 C. The materials are highly radioactive and pyrophoric, thus the FCF requires remote operation. Central to the material control and accounting system for the FCF are the balances for mass measurements. The remote operation of the balances limits direct adjustment. The radiation environment requires that removal and replacement of the balances be minimized. The uniqueness of the facility precludes historical data for design and performance assessment. To assure efficient operation of the facility, the design of the measurement control system has called for procedures which assess the performance of the balances in great detail and will support capabilities for the correction of systematic changes in the performance of the balances through software.

  6. Feasibility Study on Thermal-Hydraulic Performance of Innovative Water Reactor for Flexible Fuel Cycle (FLWR)

    SciTech Connect

    Akira, Ohnuki; Kazuyuki, Takase; Masatoshi, Kureta; Hiroyuki, Yoshida; Hidesada, Tamai; Wei, Liu; Toru, Nakatsuka; Takeharu, Misawa; Hajime, Akimoto

    2006-07-01

    R and D project to investigate thermal-hydraulic performance in tight-lattice rod bundles of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is started at Japan Atomic Energy Agency (JAEA) in collaboration with power company, reactor vendors, universities since 2002. The FLWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R and D items for the FLWR because of the tight lattice configuration. In this paper, we will show the R and D plan and summarize experimental studies. The experimental study is performed mainly using large-scale (37-rod bundle) test facility. Most important objective of the large-scale test is to resolve a fundamental subject whether the core cooling under a tight-lattice configuration is feasible. The characteristics of critical power and flow behavior are investigated under different geometrical configuration and boundary conditions. The configuration parameter is the gap between rods (FY2004) and the rod bowing (FY2005). We have confirmed the thermal-hydraulic feasibility from the experimental results. (authors)

  7. Influences of different PBL schemes on secondary eyewall formation and eyewall replacement cycle in simulated Typhoon Sinlaku (2008)

    NASA Astrophysics Data System (ADS)

    Zhang, Yutao; Jiang, Yuxin; Tan, Benkui

    2013-06-01

    The effects of different planetary boundary layer (PBL) processes on the secondary eyewall formation (SEF) and eyewall replacement cycle (ERC) in Typhoon Sinlaku (2008) are investigated by using the Weather Research and Forecasting (WRF) model with six different PBL schemes. The SEF and ERC have been successfully simulated with all the six PBL schemes and the mechanism for the SEF and ERC proposed in our previous study has been reconfirmed. It is demonstrated that both the intensification of the storm and the inward-moving outer spiral rainband contribute to the SEF. After the SEF, the associated diabatic heating enhances the secondary eyewall further, and the transfer of moist air from outer region to the primary eyewall is cut off by the secondary eyewall. In such a way, the primary eyewall dies and an ERC completes. It is found that some simulated features of the SEF and ERC, such as the time and location of the SEF and duration of the ERC, do vary from one simulation to another. In order to describe the features of the SEF and ERC quantitatively, a concentric eyewall index (CEI) is defined and a threshold of the CEI is suggested to determine the onset of the secondary eyewall. The differences in the simulated SEF and ERC are discussed and some possible causes are suggested. In addition, based on the CEI threshold and the conservation law of angular momentum, a formula to predict the location of SEF is also proposed and applied to all the six simulations. The success and failure of the formula are then discussed.

  8. Visceral Leishmaniasis on the Indian Subcontinent: Modelling the Dynamic Relationship between Vector Control Schemes and Vector Life Cycles

    PubMed Central

    2016-01-01

    Background Visceral leishmaniasis (VL) is a disease caused by two known vector-borne parasite species (Leishmania donovani, L. infantum), transmitted to man by phlebotomine sand flies (species: Phlebotomus and Lutzomyia), resulting in ≈50,000 human fatalities annually, ≈67% occurring on the Indian subcontinent. Indoor residual spraying is the current method of sand fly control in India, but alternative means of vector control, such as the treatment of livestock with systemic insecticide-based drugs, are being evaluated. We describe an individual-based, stochastic, life-stage-structured model that represents a sand fly vector population within a village in India and simulates the effects of vector control via fipronil-based drugs orally administered to cattle, which target both blood-feeding adults and larvae that feed on host feces. Principle findings Simulation results indicated efficacy of fipronil-based control schemes in reducing sand fly abundance depended on timing of drug applications relative to seasonality of the sand fly life cycle. Taking into account cost-effectiveness and logistical feasibility, two of the most efficacious treatment schemes reduced population peaks occurring from April through August by ≈90% (applications 3 times per year at 2-month intervals initiated in March) and >95% (applications 6 times per year at 2-month intervals initiated in January) relative to no control, with the cumulative number of sand fly days occurring April-August reduced by ≈83% and ≈97%, respectively, and more specifically during the summer months of peak human exposure (June-August) by ≈85% and ≈97%, respectively. Conclusions Our model should prove useful in a priori evaluation of the efficacy of fipronil-based drugs in controlling leishmaniasis on the Indian subcontinent and beyond. PMID:27537774

  9. Modeling and Simulations for the High Flux Isotope Reactor Cycle 400

    SciTech Connect

    Ilas, Germina; Chandler, David; Ade, Brian J; Sunny, Eva E; Betzler, Benjamin R; Pinkston, Daniel

    2015-03-01

    A concerted effort over the past few years has been focused on enhancing the core model for the High Flux Isotope Reactor (HFIR), as part of a comprehensive study for HFIR conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel. At this time, the core model used to perform analyses in support of HFIR operation is an MCNP model for the beginning of Cycle 400, which was documented in detail in a 2005 technical report. A HFIR core depletion model that is based on current state-of-the-art methods and nuclear data was needed to serve as reference for the design of an LEU fuel for HFIR. The recent enhancements in modeling and simulations for HFIR that are discussed in the present report include: (1) revision of the 2005 MCNP model for the beginning of Cycle 400 to improve the modeling data and assumptions as necessary based on appropriate primary reference sources HFIR drawings and reports; (2) improvement of the fuel region model, including an explicit representation for the involute fuel plate geometry that is characteristic to HFIR fuel; and (3) revision of the Monte Carlo-based depletion model for HFIR in use since 2009 but never documented in detail, with the development of a new depletion model for the HFIR explicit fuel plate representation. The new HFIR models for Cycle 400 are used to determine various metrics of relevance to reactor performance and safety assessments. The calculated metrics are compared, where possible, with measurement data from preconstruction critical experiments at HFIR, data included in the current HFIR safety analysis report, and/or data from previous calculations performed with different methods or codes. The results of the analyses show that the models presented in this report provide a robust and reliable basis for HFIR analyses.

  10. Investigation of plant control strategies for the supercritical C0{sub 2}Brayton cycle for a sodium-cooled fast reactor using the plant dynamics code.

    SciTech Connect

    Moisseytsev, A.; Sienicki, J.

    2011-04-12

    The development of a control strategy for the supercritical CO{sub 2} (S-CO{sub 2}) Brayton cycle has been extended to the investigation of alternate control strategies for a Sodium-Cooled Fast Reactor (SFR) nuclear power plant incorporating a S-CO{sub 2} Brayton cycle power converter. The SFR assumed is the 400 MWe (1000 MWt) ABR-1000 preconceptual design incorporating metallic fuel. Three alternative idealized schemes for controlling the reactor side of the plant in combination with the existing automatic control strategy for the S-CO{sub 2} Brayton cycle are explored using the ANL Plant Dynamics Code together with the SAS4A/SASSYS-1 Liquid Metal Reactor (LMR) Analysis Code System coupled together using the iterative coupling formulation previously developed and implemented into the Plant Dynamics Code. The first option assumes that the reactor side can be ideally controlled through movement of control rods and changing the speeds of both the primary and intermediate coolant system sodium pumps such that the intermediate sodium flow rate and inlet temperature to the sodium-to-CO{sub 2} heat exchanger (RHX) remain unvarying while the intermediate sodium outlet temperature changes as the load demand from the electric grid changes and the S-CO{sub 2} cycle conditions adjust according to the S-CO{sub 2} cycle control strategy. For this option, the reactor plant follows an assumed change in load demand from 100 to 0 % nominal at 5 % reduction per minute in a suitable fashion. The second option allows the reactor core power and primary and intermediate coolant system sodium pump flow rates to change autonomously in response to the strong reactivity feedbacks of the metallic fueled core and assumed constant pump torques representing unchanging output from the pump electric motors. The plant behavior to the assumed load demand reduction is surprising close to that calculated for the first option. The only negative result observed is a slight increase in the intermediate

  11. Modeling cloud microphysics using a two-moments hybrid bulk/bin scheme for use in Titan’s climate models: Application to the annual and diurnal cycles

    NASA Astrophysics Data System (ADS)

    Burgalat, J.; Rannou, P.; Cours, T.; Rivière, E. D.

    2014-03-01

    Microphysical models describe the way aerosols and clouds behave in the atmosphere. Two approaches are generally used to model these processes. While the first approach discretizes processes and aerosols size distributions on a radius grid (bin scheme), the second uses bulk parameters of the size distribution law (its mathematical moments) to represent the evolution of the particle population (moment scheme). However, with the latter approach, one needs to have an a priori knowledge of the size distributions. Moments scheme for Cloud microphysics modeling have been used and enhanced since decades for climate studies of the Earth. Most of the tools are based on Log-Normal law which are suitable for Earth, Mars or Venus. On Titan, due to the fractal structure of the aerosols, the size distributions do not follow a log-normal law. Then using a moment scheme in that case implies to define the description of the size distribution and to review the equations that are widely published in the literature. Our objective is to enable the use of a fully described microphysical model using a moment scheme within a Titan’s Global Climate Model. As a first step in this direction, we present here a moment scheme dedicated to clouds microphysics adapted for Titan’s atmosphere conditions. We perform comparisons between the two kinds of schemes (bin and moments) using an annual and a diurnal cycle, to check the validity of our moment description. The various forcing produce a time-variable cloud layer in relation with the temperature cycle. We compare the column opacities and the temperature for the two schemes, for each cycles. We also compare more detailed quantities as the opacity distribution of the cloud events at different periods of these cycles. Results show that differences between the two approaches have a small impact on the temperature (less than 1 K) and range between 1% and 10% for haze and clouds opacities. Both models behave in similar way when forced by an annual

  12. A 48-month extended fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor

    SciTech Connect

    Erighin, M. A.

    2012-07-01

    The B and W mPower{sup TM} reactor is a small, rail-shippable pressurized water reactor (PWR) with an integral once-through steam generator and an electric power output of 150 MW, which is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height, but otherwise standard, PWR assemblies with the familiar 17 x 17 fuel rod array on a 21.5 cm inter-assembly pitch. The B and W mPower core design and cycle management plan, which were performed using the Studsvik core design code suite, follow the pattern of a typical nuclear reactor fuel cycle design and analysis performed by most nuclear fuel management organizations, such as fuel vendors and utilities. However, B and W is offering a core loading and cycle management plan for four years of continuous power operations without refueling and without the hurdles of chemical shim. (authors)

  13. Reactor moderator, pressure vessel, and heat rejection system of an open-cycle gas core nuclear rocket concept

    NASA Technical Reports Server (NTRS)

    Taylor, M. F.; Whitmarsh, C. L., Jr.; Sirocky, P. J., Jr.; Iwanczyke, L. C.

    1973-01-01

    A preliminary design study of a conceptual 6000-megawatt open-cycle gas-core nuclear rocket engine system was made. The engine has a thrust of 196,600 newtons (44,200 lb) and a specific impulse of 4400 seconds. The nuclear fuel is uranium-235 and the propellant is hydrogen. Critical fuel mass was calculated for several reactor configurations. Major components of the reactor (reflector, pressure vessel, and waste heat rejection system) were considered conceptually and were sized.

  14. Development of a Scale Model for High Flux Isotope Reactor Cycle 400

    SciTech Connect

    Ilas, Dan

    2012-03-01

    The development of a comprehensive SCALE computational model for the High Flux Isotope Reactor (HFIR) is documented and discussed in this report. The SCALE model has equivalent features and functionality as the reference MCNP model for Cycle 400 that has been used extensively for HFIR safety analyses and for HFIR experiment design and analyses. Numerical comparisons of the SCALE and MCNP models for the multiplication constant, power density distribution in the fuel, and neutron fluxes at several locations in HFIR indicate excellent agreement between the results predicted with the two models. The SCALE HFIR model is presented in sufficient detail to provide the users of the model with a tool that can be easily customized for various safety analysis or experiment design requirements.

  15. Preliminary design of ultra-long cycle fast reactor employing breed-and-burn strategy

    SciTech Connect

    Tak, T. W.; Yu, H.; Kim, J. H.; Lee, D.; Kim, T. K.

    2012-07-01

    A new design of ultra-long cycle fast reactor with power rate of 1000 MWe (UCFR) has been developed based on the strategy of breed-and burn. The bottom region of the core with low enriched uranium (LEU) plays a role of igniter of the core burning and the upper natural uranium (NU) region acts as blanket for breeding. Fissile materials are bred in the blanket and the active core moves upward at a speed of 5.4 cm/year. Through the core depletion calculation using Monte Carlo code, McCARD, it is confirmed that a full power operation of 60 years without refueling is feasible. Core performance characteristics have been evaluated in terms of axial/radial power shapes, reactivity feedback coefficients, etc. This design will serve as a base model for further design study of UCFRs using LWR spent fuels in the blanket region. (authors)

  16. Analysis of closed cycle megawatt class space power systems with nuclear reactor heat sources

    NASA Technical Reports Server (NTRS)

    Juhasz, A. J.; Jones, B. I.

    1987-01-01

    The analysis and integration studies of multimegawatt nuclear power conversion systems for potential SDI applications is presented. A study is summarized which considered 3 separate types of power conversion systems for steady state power generation with a duty requirement of 1 yr at full power. The systems considered are based on the following conversion cycles: direct and indirect Brayton gas turbine, direct and indirect liquid metal Rankine, and in core thermionic. A complete mass analysis was performed for each system at power levels ranging from 1 to 25 MWe for both heat pipe and liquid droplet radiator options. In the modeling of common subsystems, reactor and shield calculations were based on multiparameter correlation and an in-house analysis for the heat rejection and other subsystems.

  17. Enhanced CANDU6: Reactor and fuel cycle options - Natural uranium and beyond

    SciTech Connect

    Ovanes, M.; Chan, P. S. W.; Mao, J.; Alderson, N.; Hopwood, J. M.

    2012-07-01

    The Enhanced CANDU 6{sup R} (ECo{sup R}) is the updated version of the well established CANDU 6 family of units incorporating improved safety characteristics designed to meet or exceed Generation III nuclear power plant expectations. The EC6 retains the excellent neutron economy and fuel cycle flexibility that are inherent in the CANDU reactor design. The reference design is based on natural uranium fuel, but the EC6 is also able to utilize additional fuel options, including the use of Recovered Uranium (RU) and Thorium based fuels, without requiring major hardware upgrades to the existing control and safety systems. This paper outlines the major changes in the EC6 core design from the existing C6 design that significantly enhance the safety characteristics and operating efficiency of the reactor. The use of RU fuel as a transparent replacement fuel for the standard 37-el NU fuel, and several RU based advanced fuel designs that give significant improvements in fuel burnup and inherent safety characteristics are also discussed in the paper. In addition, the suitability of the EC6 to use MOX and related Pu-based fuels will also be discussed. (authors)

  18. A Preliminary and Simplified Closed Brayton Cycle Modeling for a Space Reactor Application

    SciTech Connect

    Guimaraes, Lamartine Nogueira Frutuoso; Camillo, Giannino Ponchio

    2008-01-21

    The Nuclear Energy Division (ENU) of the Institute for Advanced Studies (IEAv) has started a preliminary design study for a Closed Brayton Cycle Loop (CBCL) aimed at a space reactor application. The main objectives of the study are: 1) to establish a starting concept for the CBCL components specifications, and 2) to build a demonstrative simulator of CBCL. This preliminary design study is developing the CBCL around the NOELLE 60290 turbo machine. The actual nuclear reactor study is being conducted independently. Because of that, a conventional heat source is being used for the CBCL, in this preliminary design phase. This paper describes the steady state simulator of the CBCL operating with NOELLE 60290 turbo machine. In principle, several gases are being considered as working fluid, as for instance: air, helium, nitrogen, CO{sub 2} and gas mixtures such as helium and xenon. However, for this first application pure helium will be used as working fluid. Simplified models of heat and mass transfer were developed to simulate thermal components. Future efforts will focus on implementing a graphical interface to display the thermal process variables in steady state and to keep track of the modifications being implemented at the NOELLE 60290 turbo machine in order to build the CBCL.

  19. Feasibility study of fuel cladding performance for application in ultra-long cycle fast reactor

    NASA Astrophysics Data System (ADS)

    Jung, Ju Ang; Kim, Seung Hyun; Shin, Sang Hun; Bang, In Cheol; Kim, Ji Hyun

    2013-09-01

    As a part of the research and development activities for long-life core sodium-cooled fast reactors, the cladding performance of the ultra-long cycle fast reactor (UCFR) is evaluated with two design power levels (1000 MWe and 100 MWe) and cladding peak temperatures (873 K and 923 K). The key design concept of the UCFR is that it is non-refueling during its 30-60 years of operation. This concept may require a maximum peak cladding temperature of 923 K and a cladding radiation damage of over 200 dpa (displacements per atom). Therefore, for the design of the UCFR, deformation due to thermal creep, irradiation creep, and swelling must be taken into consideration through quantitative evaluations. As candidate cladding materials for use in UCFRs, ferritic-martensitic (FM) steels, oxide dispersion strengthened (ODS) steels, and SiC-based composite materials are studied using deformation behavior modeling for a feasibility evaluation. The results of this study indicate that SiC is a potential UCFR cladding material, with the exception of irradiation creep due to high neutron fluence stemming from its long operating time of about 30-60 years.

  20. Effect of cycle changes on simultaneous biological nutrient removal in a sequencing batch reactor (SBR).

    PubMed

    Coma, M; Puig, S; Monclús, H; Balaguer, M D; Colprim, J

    2010-03-01

    The destabilization of a microbial population is sometimes hard to solve when different biological reactions are coupled in the same reactor as in sequencing batch reactors (SBRs). This paper will try to guide through practical experiences the recovery of simultaneous nitrogen and phosphorus removal in an SBR after increasing the demand of wastewater treatment by taking advantage of its flexibility. The results demonstrate that the length of phases and the optimization of influent distribution are key factors in stabilizing the system for long-term periods with high nutrient removal (88%, 93% and 99% of carbon, nitrogen and phosphorus, respectively). In order to recover a biological nutrient removal (BNR) system, different interactions such as simultaneous nitrification and denitrification and also phosphorus removal must be taken into account. As a general conclusion, it can be stated there is no such thing as a perfect SBR operation, and that much will depend on the state of the BNR system. Hence, the SBR operating strategy must be based on a dynamic cycle definition in line with process efficiency. PMID:20426270

  1. Design of a Simplified Closed Brayton Cycle for a Space Reactor Application

    SciTech Connect

    Guimaraes, Lamartine N. F.; Camillo, Giannino Ponchio; Placco, Guilherme Moreira

    2009-03-16

    The Nuclear Energy Division (ENU) of the Institute for Advanced Studies (IEAv) has started a preliminary design study for a Closed Brayton Cycle Loop (CBCL) aimed at a space reactor application. The main objectives of the study are: 1) to establish a starting concept for the CBCL components specifications, and 2) to build a demonstrative simulator of CBCL. This preliminary design study is been developed around the NOELLE 60290 turbo machine. The actual nuclear reactor study is being conducted independently. Because of that, a conventional heat source is being used for the CBCL, in this preliminary design phase. This paper describes details of the CBCL mechanical design and the steady state simulator of the CBCL operating with NOELLE 60290 turbo machine. In principle, several gases are being considered as working fluid, as for instance: air, helium, nitrogen, CO2 and gas mixtures such as helium and xenon. However, for this first application pure helium will be used as working fluid. Simplified models of heat and mass transfer were developed to simulate thermal components. A new graphical interface was developed for the simulator to display the thermal process variables in steady state and to keep track of the modifications being implemented at the NOELLE 60290 turbo machine in order to build the CBCL. A set of new results are being produced. These new results help to establish the hot and cold source geometry allowing for price estimating costs for building the actual device. These fresh new results will be presented and discussed.

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

    SciTech Connect

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

    2015-11-01

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

  3. Comparative study on nutrient removal of agricultural non-point source pollution for three filter media filling schemes in eco-soil reactors.

    PubMed

    Du, Fuyi; Xie, Qingjie; Fang, Longxiang; Su, Hang

    2016-08-01

    Nutrients (nitrogen and phosphorus) from agricultural non-point source (NPS) pollution have been increasingly recognized as a major contributor to the deterioration of water quality in recent years. The purpose of this article is to investigate the discrepancies in interception of nutrients in agricultural NPS pollution for eco-soil reactors using different filling schemes. Parallel eco-soil reactors of laboratory scale were created and filled with filter media, such as grit, zeolite, limestone, and gravel. Three filling schemes were adopted: increasing-sized filling (I-filling), decreasing-sized filling (D-filling), and blend-sized filling (B-filling). The systems were intermittent operations via simulated rainstorm runoff. The nutrient removal efficiency, biomass accumulation and vertical dissolved oxygen (DO) distribution were defined to assess the performance of eco-soil. The results showed that B-filling reactor presented an ideal DO for partial nitrification-denitrification across the eco-soil, and B-filling was the most stable in the change of bio-film accumulation trends with depth in the three fillings. Simultaneous and highest removals of NH4(+)-N (57.74-70.52%), total nitrogen (43.69-54.50%), and total phosphorus (42.50-55.00%) were obtained in the B-filling, demonstrating the efficiency of the blend filling schemes of eco-soil for oxygen transfer and biomass accumulation to cope with agricultural NPS pollution. PMID:27441855

  4. A new deflection technique applied to an existing scheme of electrostatic accelerator for high energy neutral beam injection in fusion reactor devices

    NASA Astrophysics Data System (ADS)

    Pilan, N.; Antoni, V.; De Lorenzi, A.; Chitarin, G.; Veltri, P.; Sartori, E.

    2016-02-01

    A scheme of a neutral beam injector (NBI), based on electrostatic acceleration and magneto-static deflection of negative ions, is proposed and analyzed in terms of feasibility and performance. The scheme is based on the deflection of a high energy (2 MeV) and high current (some tens of amperes) negative ion beam by a large magnetic deflector placed between the Beam Source (BS) and the neutralizer. This scheme has the potential of solving two key issues, which at present limit the applicability of a NBI to a fusion reactor: the maximum achievable acceleration voltage and the direct exposure of the BS to the flux of neutrons and radiation coming from the fusion reactor. In order to solve these two issues, a magnetic deflector is proposed to screen the BS from direct exposure to radiation and neutrons so that the voltage insulation between the electrostatic accelerator and the grounded vessel can be enhanced by using compressed SF6 instead of vacuum so that the negative ions can be accelerated at energies higher than 1 MeV. By solving the beam transport with different magnetic deflector properties, an optimum scheme has been found which is shown to be effective to guarantee both the steering effect and the beam aiming.

  5. A new deflection technique applied to an existing scheme of electrostatic accelerator for high energy neutral beam injection in fusion reactor devices.

    PubMed

    Pilan, N; Antoni, V; De Lorenzi, A; Chitarin, G; Veltri, P; Sartori, E

    2016-02-01

    A scheme of a neutral beam injector (NBI), based on electrostatic acceleration and magneto-static deflection of negative ions, is proposed and analyzed in terms of feasibility and performance. The scheme is based on the deflection of a high energy (2 MeV) and high current (some tens of amperes) negative ion beam by a large magnetic deflector placed between the Beam Source (BS) and the neutralizer. This scheme has the potential of solving two key issues, which at present limit the applicability of a NBI to a fusion reactor: the maximum achievable acceleration voltage and the direct exposure of the BS to the flux of neutrons and radiation coming from the fusion reactor. In order to solve these two issues, a magnetic deflector is proposed to screen the BS from direct exposure to radiation and neutrons so that the voltage insulation between the electrostatic accelerator and the grounded vessel can be enhanced by using compressed SF6 instead of vacuum so that the negative ions can be accelerated at energies higher than 1 MeV. By solving the beam transport with different magnetic deflector properties, an optimum scheme has been found which is shown to be effective to guarantee both the steering effect and the beam aiming. PMID:26932053

  6. A Modal Expansion Equilibrium Cycle Perturbation Method for Optimizing High Burnup Fast Reactors

    NASA Astrophysics Data System (ADS)

    Touran, Nicholas W.

    This dissertation develops a simulation tool capable of optimizing advanced nuclear reactors considering the multiobjective nature of their design. An Enhanced Equilibrium Cycle (EEC) method based on the classic equilibrium method is developed to evaluate the response of the equilibrium cycle to changes in the core design. Advances are made in the consideration of burnup-dependent cross sections and dynamic fuel performance (fission gas release, fuel growth, and bond squeeze-out) to allow accuracy in high-burnup reactors such as the Traveling Wave Reactor. EEC is accelerated for design changes near a reference state through a new modal expansion perturbation method that expands arbitrary flux perturbations on a basis of λ-eigenmodes. A code is developed to solve the 3-D, multigroup diffusion equation with an Arnoldi-based solver that determines hundreds of the reference flux harmonics and later uses these harmonics to determine expansion coefficients required to approximate the perturbed flux. The harmonics are only required for the reference state, and many substantial and localized perturbations from this state are shown to be well-approximated with efficient expressions after the reference calculation is performed. The modal expansion method is coupled to EEC to produce the later-in-time response of each design perturbation. Because the code determines the perturbed flux explicitly, a wide variety of core performance metrics may be monitored by working within a recently-developed data management system called the ARMI. Through ARMI, the response of each design perturbation may be evaluated not only for the flux and reactivity, but also for reactivity coefficients, thermal hydraulics parameters, economics, and transient performance. Considering the parameters available, an automated optimization framework is designed and implemented. A non-parametric surrogate model using the Alternating Conditional Expectation (ACE) algorithm is trained with many design

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

    NASA Astrophysics Data System (ADS)

    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

  8. Analysis of a Helium Brayton Power Cycle for a Direct-Drive Inertial Fusion Energy Power Reactor

    NASA Astrophysics Data System (ADS)

    Wagner, Scott; Gentile, Charles; Parsells, Robert; Priniski, Craig

    2008-11-01

    Presented is a thermodynamic model analysis and optimization of a helium Brayton power cycle for direct-drive inertial fusion energy (IFE) reactor. Preliminary reactor design goals include production of 2GW of thermal power and an estimated 700MW of electricity using a tertiary indirect helium Brayton cycle. A thermodynamic analysis of the proposed helium Brayton cycle is performed using baseline technology specifications and generalized thermodynamic assumptions. Analytic equations are developed using first and second law analysis. The model constraints are the turbine inlet temperature and pressure set by the reactor temperature of ˜700^oC and current turbine specifications of 7MPa, respectively. Optimization of this model is then performed using iterative numerical programming for key variables. Previous analysis shows a 51% cycle efficiency using current technology; best estimates of near-term technology increase the cycle efficiency to 64%. Results will be presented. R. Schleicher, A. R. Raffray, C. P. Wong, ``An Assessment of the Brayton Cycle for High Performance Power Plant,'' Fusion Technology, 39 (2), 823-827, March 2001.

  9. Parametric analyses of single-zone thorium-fueled molten salt reactor fuel cycle options

    SciTech Connect

    Powers, J.J.; Worrall, A.; Gehin, J.C.; Harrison, T.J.; Sunny, E.E.

    2013-07-01

    Analyses of fuel cycle options based on thorium-fueled Molten Salt Reactors (MSRs) have been performed in support of fuel cycle screening and evaluation activities for the United States Department of Energy. The MSR options considered are based on thermal spectrum MSRs with 3 different separations levels: full recycling, limited recycling, and 'once-through' operation without active separations. A single-fluid, single-zone 2250 MWth (1000 MWe) MSR concept consisting of a fuel-bearing molten salt with graphite moderator and reflectors was used as the basis for this study. Radiation transport and isotopic depletion calculations were performed using SCALE 6.1 with ENDF/B-VII nuclear data. New methodology developed at Oak Ridge National Laboratory (ORNL) enables MSR analysis using SCALE, modeling material feed and removal by taking user-specified parameters and performing multiple SCALE/TRITON simulations to determine the resulting equilibrium operating conditions. Parametric analyses examined the sensitivity of the performance of a thorium MSR to variations in the separations efficiency for protactinium and fission products. Results indicate that self-sustained operation is possible with full or limited recycling but once-through operation would require an external neutron source. (authors)

  10. Management of transuranics using the Integral Fast Reactor (IFR) fuel cycle

    SciTech Connect

    Wade, D.C.

    1994-01-01

    The 50 years of activities following the discovery of self-sustaining fission chains have given rise to a buildup of roughly 900 tons of manmade transuranics. Of the total, about 260 tons of Pu{sup 239} were generated for use in weapons while the remainder were generated as a byproduct of electrical power produced worldwide by the commercial thermal nuclear power industry. What is to be done with these actinides? The options for disposition include interminable storage, burial, or recycle for use. The pros and cons of each option are being vigorously debated regarding the impact upon the issues of human and ecological risk -- both current and future; weapons proliferation potential -- both current and future; and total life cycle benefits and costs. As to the options for utilization, commercial uses for actinides (uranium and transuranics) are of limited diversity. The actinides have in the past and will in the future find application in large scale mostly by virtue of their ability to release energy through fission, and here their utility is unmatched -- whether the application be in commercial electricity generation or in armaments. The integral Fast Reactor (IFR) fuel cycle offers a number of features for management of the current and future burden of manmade transuranic materials and for capturing the energy content of the U{sup 238}. These features are discussed here.

  11. A Burst Mode, Ultrahigh Temperature UF4 Vapor Core Reactor Rankine Cycle Space Power System Concept

    NASA Technical Reports Server (NTRS)

    Dugan, E. T.; Kahook, S. D.; Diaz, N. J.

    1996-01-01

    Static and dynamic neutronic analyses have been performed on an innovative burst mode (100's of MW output for a few thousand seconds) Ulvahigh Temperature Vapor Core Reactor (UTVR) space nuclear power system. The NVTR employs multiple, neutronically-coupled fissioning cores and operates on a direct, closed Rankine cycle using a disk Magnetohydrodynamic (MHD) generater for energy conversion. The UTVR includes two types of fissioning core regions: (1) the central Ultrahigh Temperature Vapor Core (UTVC) which contains a vapor mixture of highly enriched UF4 fuel and a metal fluoride working fluid and (2) the UF4 boiler column cores located in the BeO moderator/reflector region. The gaseous nature of the fuel the fact that the fuel is circulating, the multiple coupled fissioning cores, and the use of a two phase fissioning fuel lead to unique static and dynamic neutronic characteristics. Static neutronic analysis was conducted using two-dimensional S sub n, transport theory calculations and three-dimensional Monte Carlo transport theory calculations. Circulating-fuel, coupled-core point reactor kinetics equations were used for analyzing the dynamic behavior of the UTVR. In addition to including reactivity feedback phenomena associated with the individual fissioning cores, the effects of core-to-core neutronic and mass flow coupling between the UTVC and the surrounding boiler cores were also included in the dynamic model The dynamic analysis of the UTVR reveals the existence of some very effectlve inherent reactivity feedback effects that are capable of quickly stabilizing this system, within a few seconds, even when large positive reactivity insertions are imposed. If the UTVC vapor fuel density feedback is suppressed, the UTVR is still inherently stable because of the boiler core liquid-fuel volume feedback; in contrast, suppression of the vapor fuel density feedback in 'conventional" gas core cavity reactors causes them to become inherently unstable. Due to the

  12. An extended conventional fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor

    SciTech Connect

    Scarangella, M. J.

    2012-07-01

    The B and W mPower{sup TM} reactor is a small pressurized water reactor (PWR) with an integral once-through steam generator and a thermal output of about 500 MW; it is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height PWR assemblies with the familiar 17 x 17 fuel rod array. The Babcock and Wilcox Company (B and W) is offering a core loading and cycle management plan for a four-year cycle based on its presumed attractiveness to potential customers. This option is a once-through fuel cycle in which the entire core is discharged and replaced after four years. In addition, a conventional fuel utilization strategy, employing a periodic partial reload and shuffle, was developed as an alternative to the four-year once-through fuel cycle. This study, which was performed using the Studsvik core design code suite, is a typical multi-cycle projection analysis of the type performed by most fuel management organizations such as fuel vendors and utilities. In the industry, the results of such projections are used by the financial arms of these organizations to assist in making long-term decisions. In the case of the B and W mPower reactor, this analysis demonstrates flexibility for customers who consider the once-through fuel cycle unacceptable from a fuel utilization standpoint. As expected, when compared to the once-through concept, reloads of the B and W mPower reactor will achieve higher batch average discharge exposure, will have adequate shut-down margin, and will have a relatively flat hot excess reactivity trend at the expense of slightly increased peaking. (authors)

  13. Tokamak burn cycle study: a data base for comparing long pulse and steady-state power reactors

    SciTech Connect

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K. Jr.; Hassanein, A.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1983-11-01

    Several distinct operating modes (conventional ohmic, noninductive steady state, internal transformer, etc.) have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics (current drive efficiency) and engineering (superior materials) which will help achieve these goals for different burn cycles.

  14. Data Reconciliation in the Steam-Turbine Cycle of a Boiling Water Reactor

    SciTech Connect

    Sunde, Svein; Berg, Oivind; Dahlberg, Lennart; Fridqvist, Nils-Olof

    2003-08-15

    A mathematical model for a boiling water reactor steam-turbine cycle was assembled by means of a configurable, steady-state modeling tool TEMPO. The model was connected to live plant data and intermittently fitted to these by minimization of a weighted least-squares object function. The improvement in precision achieved by this reconciliation was assessed from quantities calculated from the model equations linearized around the minimum and from Monte Carlo simulations. It was found that the inclusion of the flow-passing characteristics of the turbines in the model equations significantly improved the precision as compared to simple mass and energy balances, whereas heat transfer calculations in feedwater heaters did not. Under the assumption of linear model equations, the quality of the fit can also be expressed as a goodness-of-fit Q. Typical values for Q were in the order of 0.9. For a validated model Q may be used as a fault detection indicator, and Q dropped to very low values in known cases of disagreement between the model and the plant state. The sensitivity of Q toward measurement faults is discussed in relation to redundancy. The results of the linearized theory and Monte Carlo simulations differed somewhat, and if a more accurate analysis is required, this is better based on the latter. In practical application of the presently employed techniques, however, assessment of uncertainties in raw data is an important prerequisite.

  15. The thermodynamics of pyrochemical processes for liquid metal reactor fuel cycles

    SciTech Connect

    Johnson, I.

    1987-01-01

    The thermodynamic basis for pyrochemical processes for the recovery and purification of fuel for the liquid metal reactor fuel cycle is described. These processes involve the transport of the uranium and plutonium from one liquid alloy to another through a molten salt. The processes discussed use liquid alloys of cadmium, zinc, and magnesium and molten chloride salts. The oxidation-reduction steps are done either chemically by the use of an auxiliary redox couple or electrochemically by the use of an external electrical supply. The same basic thermodynamics apply to both the salt transport and the electrotransport processes. Large deviations from ideal solution behavior of the actinides and lanthanides in the liquid alloys have a major influence on the solubilities and the performance of both the salt transport and electrotransport processes. Separation of plutonium and uranium from each other and decontamination from the more noble fission product elements can be achieved using both transport processes. The thermodynamic analysis is used to make process design computations for different process conditions.

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

    SciTech Connect

    Moisseytsev, A.; Sienicki, J. J.

    2009-07-01

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

  17. Use of RELAP5-3D for Dynamic Analysis of a Closed-Loop Brayton Cycle Coupled To a Nuclear Reactor

    SciTech Connect

    McCann, Larry D.

    2007-01-30

    This paper describes results of a dynamic system model for a pair of closed Brayton-cycle (CBC) loops running in parallel that are connected to a nuclear gas reactor. The model assumes direct coupling between the reactor and the Brayton-cycle loops. The RELAP5-3D (version 2.4.1) computer program was used to perform the analysis. Few reactors have ever been coupled to closed Brayton-cycle systems. As such their behavior under dynamically varying loads, startup and shut down conditions, and requirements for safe and autonomous operation are largely unknown. The model described in this paper represents the reactor, turbine, compressor, recuperator, heat rejection system and alternator. The initial results of the model indicate stable operation of the reactor-driven Brayton-cycle system. However, for analysts with mostly pressurized water reactor experience, the Brayton cycle loops coupled to a gas-cooled reactor also indicate some counter-intuitive behavior for the complete coupled system. This model has provided crucial information in evaluating the reactor design and would have been further developed for use in developing procedures for safe start up, shut down, safe-standby, and other autonomous operating modes had the plant development cycle been completed.

  18. Use of RELAP5-3D for Dynamic Analysis of a Closed-Loop Brayton Cycle Coupled To a Nuclear Reactor

    NASA Astrophysics Data System (ADS)

    McCann, Larry D.

    2007-01-01

    This paper describes results of a dynamic system model for a pair of closed Brayton-cycle (CBC) loops running in parallel that are connected to a nuclear gas reactor. The model assumes direct coupling between the reactor and the Brayton-cycle loops. The RELAP5-3D (version 2.4.1) computer program was used to perform the analysis. Few reactors have ever been coupled to closed Brayton-cycle systems. As such their behavior under dynamically varying loads, startup and shut down conditions, and requirements for safe and autonomous operation are largely unknown. The model described in this paper represents the reactor, turbine, compressor, recuperator, heat rejection system and alternator. The initial results of the model indicate stable operation of the reactor-driven Brayton-cycle system. However, for analysts with mostly pressurized water reactor experience, the Brayton cycle loops coupled to a gas-cooled reactor also indicate some counter-intuitive behavior for the complete coupled system. This model has provided crucial information in evaluating the reactor design and would have been further developed for use in developing procedures for safe start up, shut down, safe-standby, and other autonomous operating modes had the plant development cycle been completed.

  19. Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities. Sections 1-14

    SciTech Connect

    Schneider, K.J.

    1982-09-01

    The Fuel Cycle Risk Assessment Program was initiated to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. This report, the first from the program, defines and describes fuel cycle elements that are being considered in the program. One type of facility (and in some cases two) is described that is representative of each element of the fuel cycle. The descriptions are based on real industrial-scale facilities that are current state-of-the-art, or on conceptual facilities where none now exist. Each representative fuel cycle facility is assumed to be located on the appropriate one of four hypothetical but representative sites described. The fuel cycles considered are for Light Water Reactors with once-through flow of spent fuel, and with plutonium and uranium recycle. Representative facilities for the following fuel cycle elements are described for uranium (or uranium plus plutonium where appropriate): mining, milling, conversion, enrichment, fuel fabrication, mixed-oxide fuel refabrication, fuel reprocessing, spent fuel storage, high-level waste storage, transuranic waste storage, spent fuel and high-level and transuranic waste disposal, low-level and intermediate-level waste disposal, and transportation. For each representative facility the description includes: mainline process, effluent processing and waste management, facility and hardware description, safety-related information and potential alternative concepts for that fuel cycle element. The emphasis of the descriptive material is on safety-related information. This includes: operating and maintenance requirements, input/output of major materials, identification and inventories of hazardous materials (particularly radioactive materials), unit operations involved, potential accident driving forces, containment and shielding, and degree of hands-on operation.

  20. Closed Brayton Cycle power system with a high temperature pellet bed reactor heat source for NEP applications

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.; El-Genk, Mohamed S.; Harper, William B., Jr.

    1992-01-01

    Capitalizing on past and future development of high temperature gas reactor (HTGR) technology, a low mass 15 MWe closed gas turbine cycle power system using a pellet bed reactor heating helium working fluid is proposed for Nuclear Electric Propulsion (NEP) applications. Although the design of this directly coupled system architecture, comprising the reactor/power system/space radiator subsystems, is presented in conceptual form, sufficient detail is included to permit an assessment of overall system performance and mass. Furthermore, an attempt is made to show how tailoring of the main subsystem design characteristics can be utilized to achieve synergistic system level advantages that can lead to improved reliability and enhanced system life while reducing the number of parasitic load driven peripheral subsystems.

  1. Design studies of the sulfur trioxide decomposition reactor for the sulfur-cycle hydrogen-production process

    SciTech Connect

    Lin, S.S.; Flaherty, R.

    1982-01-01

    The Sulfur Cycle is a two-step hybrid electrochemical/thermochemical process for decomposing water into hydrogen and oxygen. Integration of a complex chemical process with a solar heat source poses unique challenges with regard to process and equipment design. The conceptual design for a developmental test unit demonstrating the sulfur cycle was prepared in 1980. The test unit design is compatible with the power level of a large parabolic solar collector. One of the key components in the process is the sulfur trioxide decomposition reactor. The design studies of the sulfur trioxide decomposition reactor encompassing the thermodynamics, reaction kinetics, heat transfer, and mechanical considerations, are described along with a brief description of the test unit.

  2. Renovation of CPF (Chemical Processing Facility) for Development of Advanced Fast Reactor Fuel Cycle System

    SciTech Connect

    Shinichi Aose; Takafumi Kitajima; Kouji Ogasawara; Kazunori Nomura; Shigehiko Miyachi; Yoshiaki Ichige; Tadahiro Shinozaki; Shinichi Ohuchi

    2008-01-15

    CPF (Chemical Processing Facility) was constructed at Nuclear Fuel Cycle Engineering Laboratories of JAEA (Japan Atomic Energy Agency) in 1980 as a basic research field where spent fuel pins from fast reactor (FR) and high level liquid waste can be dealt with. The renovation consists of remodeling of the CA-3 cell and the laboratory A, installation of globe boxes, hoods and analytical equipments to the laboratory C and the analytical laboratory. Also maintenance equipments in the CA-5 cell which had been out of order were repaired. The CA-3 cell is the main cell in which important equipments such as a dissolver, a clarifier and extractors are installed for carrying out the hot test using the irradiated FR fuel. Since the CPF had specialized originally in the research function for the Purex process, it was desired to execute the research and development of such new, various reprocessing processes. Formerly, equipments were arranged in wide space and connected with not only each other but also with utility supply system mainly by fixed stainless steel pipes. It caused shortage of operation space in flexibility for basic experimental study. Old equipments in the CA-3 cell including vessels and pipes were removed after successful decontamination, and new equipments were installed conformably to the new design. For the purpose of easy installation and rearranging the experimental equipments, equipments are basically connected by flexible pipes. Since dissolver is able to be easily replaced, various dissolution experiments is conducted. Insoluble residue generated by dissolution of spent fuel is clarified by centrifugal. This small apparatus is effective to space-saving. Mini mixer settlers or centrifugal contactors are put on to the prescribed limited space in front of the backside wall. Fresh reagents such as solvent, scrubbing and stripping solution are continuously fed from the laboratory A to the extractor by the reagent supply system with semi-automatic observation

  3. Effects of Aeration Cycles on Nitrifying Bacterial Populations and Nitrogen Removal in Intermittently Aerated Reactors

    PubMed Central

    Mota, Cesar; Head, Melanie A.; Ridenoure, Jennifer A.; Cheng, Jay J.; de los Reyes, Francis L.

    2005-01-01

    The effects of the lengths of aeration and nonaeration periods on nitrogen removal and the nitrifying bacterial community structure were assessed in intermittently aerated (IA) reactors treating digested swine wastewater. Five IA reactors were operated in parallel with different aeration-to-nonaeration time ratios (ANA). Populations of ammonia-oxidizing bacteria (AOB) and nitrite-oxidizing bacteria (NOB) were monitored using 16S rRNA slot blot hybridizations. AOB species diversity was assessed using amoA gene denaturant gradient gel electrophoresis. Nitrosomonas and Nitrosococcus mobilis were the dominant AOB and Nitrospira spp. were the dominant NOB in all reactors, although Nitrosospira and Nitrobacter were also detected at lower levels. Reactors operated with the shortest aeration time (30 min) showed the highest Nitrosospira rRNA levels, and reactors operated with the longest anoxic periods (3 and 4 h) showed the lowest levels of Nitrobacter, compared to the other reactors. Nitrosomonas sp. strain Nm107 was detected in all reactors, regardless of the reactor's performance. Close relatives of Nitrosomonas europaea, Nitrosomonas sp. strain ENI-11, and Nitrosospira multiformis were occasionally detected in all reactors. Biomass fractions of AOB and effluent ammonia concentrations were not significantly different among the reactors. NOB were more sensitive than AOB to long nonaeration periods, as nitrite accumulation and lower total NOB rRNA levels were observed for an ANA of 1 h:4 h. The reactor with the longest nonaeration time of 4 h performed partial nitrification, followed by denitrification via nitrite, whereas the other reactors removed nitrogen through traditional nitrification and denitrification via nitrate. Superior ammonia removal efficiencies were not associated with levels of specific AOB species or with higher AOB species diversity. PMID:16332848

  4. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    SciTech Connect

    Lindley, Benjamin A.; Parks, Geoffrey T.; Franceschini, Fausto

    2013-07-01

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

  5. Steam Generator Component Model in a Combined Cycle of Power Conversion Unit for Very High Temperature Gas-Cooled Reactor

    SciTech Connect

    Oh, Chang H; Han, James; Barner, Robert; Sherman, Steven R

    2007-06-01

    The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP), Very High Temperature Gas-Cooled Reactor (VHTR) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. A combined cycle is considered as one of the power conversion units to be coupled to the very high-temperature gas-cooled reactor (VHTR). The combined cycle configuration consists of a Brayton top cycle coupled to a Rankine bottoming cycle by means of a steam generator. A detailed sizing and pressure drop model of a steam generator is not available in the HYSYS processes code. Therefore a four region model was developed for implementation into HYSYS. The focus of this study was the validation of a HYSYS steam generator model of two phase flow correlations. The correlations calculated the size and heat exchange of the steam generator. To assess the model, those calculations were input into a RELAP5 model and its results were compared with HYSYS results. The comparison showed many differences in parameters such as the heat transfer coefficients and revealed the different methods used by the codes. Despite differences in approach, the overall results of heat transfer were in good agreement.

  6. Importance of the (n,gamma) Cm-247 Evaluation on Neutron Emission in Fast Reactor Fuel Cycle Analysis

    SciTech Connect

    Benoit Forget; Mehdi Asgari; Rodolfo M. Ferrer

    2007-11-01

    As part of the GNEP program, it is envisioned to build a fast reactor for the transmutation of minor actinides. The spent nuclear fuel from the current fleet of light water reactors would be recycled, the current baseline is the UREX+1a process, and would act as a feed for the fast reactor. As the fuel is irradiated in a fast reactor a certain quantity of minor actinides would thus build up in the fuel stream creating possible concerns with the neutron emission of these minor actinides for fuel transportation, handling and fabrication. Past neutronic analyses had not tracked minor actinides above Cm-246 in the transmutation chain, because of the small influence on the overall reactor performance and cycle parameters. However, when trying to quantify the neutron emission from the recycled fuel with high minor actinide content, these higher isotopes play an essential role and should be included in the analysis. In this paper, the influence of tracking these minor actinides on the calculated neutron emission is presented. Also presented is the particular influence of choosing a different evaluated cross section data set to represent the minor actinides above Cm-246. The first representation uses the cross-sections provided by MC2-2 for all isotopes, while the second representation uses infinitely diluted ENDF/BVII.0 cross-sections for Cm-247 to Cf-252 and MC2-2 for all other isotopes.

  7. Reactor physics studies for the Advanced Fuel Cycle Initiative (AFCI) Reactor-Accelerator Coupling Experiments (RACE) Project

    NASA Astrophysics Data System (ADS)

    Stankovskiy, Evgeny Yuryevich

    In the recently completed RACE Project of the AFCI, accelerator-driven subcritical systems (ADS) experiments were conducted to develop technology of coupling accelerators to nuclear reactors. In these experiments electron accelerators induced photon-neutron reactions in heavy-metal targets to initiate fission reactions in ADS. Although the Idaho State University (ISU) RACE ADS was constructed only to develop measurement techniques for advanced experiments, many reactor kinetics experiments were conducted there. In the research reported in this dissertation, a method was developed to calculate kinetics parameters for measurement and calculation of the reactivity of ADS, a safety parameter that is necessary for control and monitoring of power production. Reactivity is measured in units of fraction of delayed versus prompt neutron from fission, a quantity that cannot be directly measured in far-subcritical reactors such as the ISU RACE configuration. A new technique is reported herein to calculate it accurately and to predict kinetic behavior of a far-subcritical ADS. Experiments conducted at ISU are first described and experimental data are presented before development of the kinetic theory used in the new computational method. Because of the complexity of the ISU ADS, the Monte-Carlo method as applied in the MCNP code is most suitable for modeling reactor kinetics. However, the standard method of calculating the delayed neutron fraction produces inaccurate values. A new method was developed and used herein to evaluate actual experiments. An advantage of this method is that its efficiency is independent of the fission yield of delayed neutrons, which makes it suitable for fuel with a minor actinide component (e.g. transmutation fuels). The implementation of this method is based on a correlated sampling technique which allows the accurate evaluation of delayed and prompt neutrons. The validity of the obtained results is indicated by good agreement between experimental

  8. ENERGY EFFICIENCY LIMITS FOR A RECUPERATIVE BAYONET SULFURIC ACID DECOMPOSITION REACTOR FOR SULFUR CYCLE THERMOCHEMICAL HYDROGEN PRODUCTION

    SciTech Connect

    Gorensek, M.; Edwards, T.

    2009-06-11

    A recuperative bayonet reactor design for the high-temperature sulfuric acid decomposition step in sulfur-based thermochemical hydrogen cycles was evaluated using pinch analysis in conjunction with statistical methods. The objective was to establish the minimum energy requirement. Taking hydrogen production via alkaline electrolysis with nuclear power as the benchmark, the acid decomposition step can consume no more than 450 kJ/mol SO{sub 2} for sulfur cycles to be competitive. The lowest value of the minimum heating target, 320.9 kJ/mol SO{sub 2}, was found at the highest pressure (90 bar) and peak process temperature (900 C) considered, and at a feed concentration of 42.5 mol% H{sub 2}SO{sub 4}. This should be low enough for a practical water-splitting process, even including the additional energy required to concentrate the acid feed. Lower temperatures consistently gave higher minimum heating targets. The lowest peak process temperature that could meet the 450-kJ/mol SO{sub 2} benchmark was 750 C. If the decomposition reactor were to be heated indirectly by an advanced gas-cooled reactor heat source (50 C temperature difference between primary and secondary coolants, 25 C minimum temperature difference between the secondary coolant and the process), then sulfur cycles using this concept could be competitive with alkaline electrolysis provided the primary heat source temperature is at least 825 C. The bayonet design will not be practical if the (primary heat source) reactor outlet temperature is below 825 C.

  9. Role of fast reactor and its cycle to reduce nuclear waste burden

    SciTech Connect

    Arie, Kazuo; Oomori, Takashi; Okita, Takeshi; Kawashima, Masatoshi; Kotake, Shoji; Fuji-ie, Yoichi

    2013-07-01

    The role of the metal fuel fast reactor with recycling of actinides and the five long-lived fission products based on the concept of the Self-Consistent Nuclear Energy System has been examined by evaluating the reduction of nuclear wastes during the transition period to this reactor system. The evaluation was done in comparison to an LWR once-through case and a conventional actinide recycling oxide fast reactor. As a result, it is quantitatively clarified that a metal fuel fast reactor with actinide and the five long-lived fission products (I{sup 129}, Tc{sup 99}, Zr{sup 93}, Cs{sup 135} and Sn{sup 126}) recycling could play a significant role in reducing the nuclear waste burden including the current LWR wastes. This can be achieved by using a fast neutron spectrum reactor enhanced with metal fuel that brings high capability as a 'waste burner'. (authors)

  10. Increasing the reliability of the shutdown of 500 - 750-kV overhead lines equipped with shunt reactors in an unsuccessful three-phase automatic repeated closure cycle

    SciTech Connect

    Kuz'micheva, K. I.; Merzlyakov, A. S.; Fokin, G. G.

    2013-05-15

    The reasons for circuit-breaker failures during repeated disconnection of 500 - 750 kV overhead lines with shunt reactors in a cycle of unsuccessful three-phase automatic reconnection (TARC) are analyzed. Recommendations are made for increasing the operating reliability of power transmission lines with shunt reactors when there is unsuccessful reconnection.

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

    SciTech Connect

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

    2013-03-01

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

  12. Fuel cycles and envisioned roles of fast neutron reactors and hybrids

    NASA Astrophysics Data System (ADS)

    Salvatores, Massimo

    2012-06-01

    Future innovative nuclear fuel cycles will require insuring sustainability in terms of safe operation, optimal use of resources, radioactive waste minimization and reduced risk of proliferation. The present paper introduces some basic notions and fundamental fuel cycle strategies. The simulation approach needed to evaluate the impact of the different fuel cycle alternatives will also be shortly discussed.

  13. Fuel cycles and envisioned roles of fast neutron reactors and hybrids

    SciTech Connect

    Salvatores, Massimo

    2012-06-19

    Future innovative nuclear fuel cycles will require insuring sustainability in terms of safe operation, optimal use of resources, radioactive waste minimization and reduced risk of proliferation. The present paper introduces some basic notions and fundamental fuel cycle strategies. The simulation approach needed to evaluate the impact of the different fuel cycle alternatives will also be shortly discussed.

  14. Record of Cycling Operation of the Natural Nuclear Reactor in the Oklo/Okelobondo Area in Gabon

    NASA Astrophysics Data System (ADS)

    Meshik, A. P.; Hohenberg, C. M.; Pravdivtseva, O. V.

    2004-10-01

    Using selective laser extraction technique combined with sensitive ion-counting mass spectrometry, we have analyzed the isotopic structure of fission noble gases in U-free La-Ce-Sr-Ca aluminous hydroxy phosphate associated with the 2 billion yr old Oklo natural nuclear reactor. In addition to elevated abundances of fission-produced Zr, Ce, and Sr, we discovered high (up to 0.03 cm3 STP/g) concentrations of fission Xe and Kr, the largest ever observed in any natural material. The specific isotopic structure of xenon in this mineral defines a cycling operation for the reactor with 30-min active pulses separated by 2.5h dormant periods. Thus, nature not only created conditions for self-sustained nuclear chain reactions, but also provided clues on how to retain nuclear wastes, including fission Xe and Kr, and prevent uncontrolled runaway chain reaction.

  15. Record of cycling operation of the natural nuclear reactor in the Oklo/Okelobondo area in Gabon.

    PubMed

    Meshik, A P; Hohenberg, C M; Pravdivtseva, O V

    2004-10-29

    Using selective laser extraction technique combined with sensitive ion-counting mass spectrometry, we have analyzed the isotopic structure of fission noble gases in U-free La-Ce-Sr-Ca aluminous hydroxy phosphate associated with the 2 billion yr old Oklo natural nuclear reactor. In addition to elevated abundances of fission-produced Zr, Ce, and Sr, we discovered high (up to 0.03 cm(3) STP/g) concentrations of fission Xe and Kr, the largest ever observed in any natural material. The specific isotopic structure of xenon in this mineral defines a cycling operation for the reactor with 30-min active pulses separated by 2.5 h dormant periods. Thus, nature not only created conditions for self-sustained nuclear chain reactions, but also provided clues on how to retain nuclear wastes, including fission Xe and Kr, and prevent uncontrolled runaway chain reaction. PMID:15525157

  16. Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

    SciTech Connect

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal O. Pasamehmetoglu

    2012-04-01

    The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather than graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water

  17. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 5 - North Anna Unit 1 Cycle 5

    SciTech Connect

    Bowman, S.M.

    1993-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor (AFR) criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark AFR criticality analysis methods using selected critical configurations from commercial pressurized-water reactors (PWR). The analysis methodology selected for all calculations reported herein was the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and to provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of one reactor critical configuration for North Anna Unit 1 Cycle 5. This unit and cycle were chosen for a previous analysis using a different methodology because detailed isotopics from multidimensional reactor calculations were available from the Virginia Power Company. These data permitted comparison of criticality calculations directly using the utility-calculated isotopics to those using the isotopics generated by the SCALE-4 SAS2H

  18. Present and future status of thermochemical cycles applied to fusion energy sources

    SciTech Connect

    Booth, L.A.; Cox, K.E.; Krakowski, R.A.; Pendergrass, J.H.

    1980-01-01

    This paper reviews the status of current research on thermochemical hydrogen production cycles and identifies the needs for advanced cycles and materials research. The Los Alamos Scientific Laboratory (LASL) bismuth sulfate thermochemical cycle is characterized, and fusion reactor blanket concepts for both inertial and magnetic confinement schemes are presented as thermal energy sources for process heat applications.

  19. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  20. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  1. Safeguards and Non-proliferation Issues as Related to Advanced Fuel Cycle and Advanced Fast Reactor Development with Processing of Reactor Fuel

    SciTech Connect

    Rahmat Aryaeinejad; Jerry D. Cole; Mark W. Drigert; Dee E. Vaden

    2006-10-01

    The goal of this work is to establish basic data and techniques to enable safeguards appropriate to a new generation of nuclear power systems that will be based on fast spectrum reactors and mixed actinide fuels containing significant quantities of "minor" actinides, possibly due to reprocessing, and determination of what new radiation signatures and parameters need to be considered. The research effort focuses on several problems associated with the use of fuel having significantly different actinide inventories that current practice and on the development of innovative techniques using new radiation signatures and other parameters useful for safeguards and monitoring. In addition, the development of new distinctive radiation signatures as an aid in controlling proliferation of nuclear materials has parallel applications to support Gen-IV and current advanced fuel cycle initiative (AFCI) goals as well as the anticipated Global Nuclear Energy Partnership (GNEP).

  2. Translational activities of EF-Tu [G222D] which cannot be reconciled with the classical scheme of the polypeptide chain elongation cycle.

    PubMed

    Talens, A; Boon, K; Kraal, B; Bosch, L

    1996-08-23

    We have developed a cell-free system of E. coli that enables us to study the in vitro translation of natural mRNA mediated by wild-type or mutant EF-Tu. Various mutant EF-Tu species have been analyzed, one of which, EF-Tu [G222D], appeared to be virtually unable to mediate the translation of natural mRNA. Since this mutant factor is able to participate in translation in vivo by suppressing nonsense and frameshift mutations in cooperation with EF-Tu [A375T], a revision of the generally accepted scheme of the elongation cycle has been proposed (Bosch, L., Vijgenboom, E., & Zeef, L.A.H., 1996, Biochemistry 36). PMID:8780718

  3. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 4-Three Mile Island Unit 1 Cycle 5

    SciTech Connect

    DeHart, M.D.

    1995-01-01

    The requirements of ANSI/ANS-8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original ''fresh'' composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using relevant and well-documented critical configurations from commercial pressurized water reactors. The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Isotopic densities for spent fuel assemblies in the core were calculated using the SCALE-4 SAS2H analytical sequence. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code family was used to extract the necessary isotopic densities from SAS2H results and to provide the data in the format required for SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) for the critical configuration. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all calculations. This volume of the report documents a reactor critical calculation for GPU Nuclear Corporation's Three Mile Island Unit 1 (TMI-1) during hot, zero-power startup testing for the beginning of cycle 5. This unit and cycle were selected because of their relevance in spent fuel benchmark applications: (1) cycle 5 startup occurred after an especially long downtime of 6.6 years; and (2) the core consisted primarily (75%) of burned fuel, with

  4. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  5. Materials considerations for the coupling of thermochemical hydrogen cycles to tandem mirror reactors

    SciTech Connect

    Krikorian, O.H.

    1980-10-10

    Candidate materials are discussed and initial choices made for the critical elements in a liquid Li-Na Cauldron Tandem Mirror blanket and the General Atomic Sulfur-Iodine Cycle for thermochemical hydrogen production. V and Ti alloys provide low neutron activation, good radiation damage resistance, and good chemical compatibility for the Cauldron design. Aluminide coated In-800H and siliconized SiC are materials choices for heat exchanger components in the thermochemical cycle interface.

  6. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  7. Implications of the DD fuel cycle on tokamak reactor technology considerations

    SciTech Connect

    Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Evans, K. Jr.; Jung, J.; Mattas, R.; Misra, B.; Smith, D.L.; Stevens, H.C.; Turner, L.R.

    1980-01-01

    The initial results of a study to develop a conceptual design of a DD tokamak reactor are reported in this paper. A comprehensive steady-state power and particle balance code has been used to survey a variety of design points. A reference set of parameters (R/sub 0/ = 8.6 m, B/sub max/ = 14 T, ..beta.. = 11%, P/sub th/ = 2500 to 2700 MW) has been selected for further study. The concepts of rf driven currents and a limiter-vacuum system have been examined and appear to offer promise. However, the penalty of rf current drive, in terms of recirculating power, is relatively larger for DD than for DT reactors. Blanket engineering studies have emphasized the optimization of energy multiplication and inboard blanket thickness.

  8. Space reactor/Stirling cycle systems for high power Lunar applications

    SciTech Connect

    Schmitz, P.D.; Mason, L.S.

    1994-09-01

    NASA`s Space Exploration Initiative (SEI) has proposed the use of high power nuclear power systems on the lunar surface as a necessary alternative to solar power. Because of the long lunar night ({approximately} 14 earth days) solar powered systems with the requisite energy storage in the form of regenerative fuel cells or batteries becomes prohibitively heavy at high power levels ({approximately} 100 kWe). At these high power levels nuclear power systems become an enabling technology for variety of missions. One way of producing power on the lunar surface is with an SP-100 class reactor coupled with Stirling power converters. In this study, analysis and characterization of the SP-100 class reactor coupled with Free Piston Stirling Power Conversion (FPSPC) system will be performed. Comparison of results with previous studies of other systems, particularly Brayton and Thermionic, are made.

  9. Space reactor/Stirling cycle systems for high power lunar applications

    NASA Technical Reports Server (NTRS)

    Schmitz, Paul C.; Mason, Lee S.

    1991-01-01

    An analysis is performed to mathematically model a 550 kWe lunar base power supply which uses a SP-100 reactor coupled with Stirling converters. The reactor is placed in an excavation to keep activated coolant in the hole and to allow maintance of the components outside the hole. Two technology levels are considered. They are 1050 and 1300 K heater head Stirling converts. It is found that for a 1050 K converter the total mass which provided 1000 volts dc at 250 m is 14,366 kg while the 1300 K system mass is 12,104 kg. The radiation area of the 1050 and 1300 K systems are 641 and 356 sq m respectively. Comparisons are made with Brayton and thermionic systems with both near term and advanced technology considered.

  10. Space reactor/Stirling cycle systems for high power lunar application

    NASA Technical Reports Server (NTRS)

    Schmitz, Paul C.; Mason, Lee S.

    1991-01-01

    An analysis is performed to mathematically model a 550 kWe lunar base power supply which uses a SP-100 reactor coupled with Stirling converters. The reactor is placed in an excavation to keep activated coolant in the hole and to allow maintenance of the components outside the hole. Two technology levels are considered. They are 1050 and 1300 K heater head Stirling converts. It is found that for a 1050 K converter the total mass which provided 1000 volts DC at 250 m is 14,366 kg while the 1300 K system mass is 12,104 kg. The radiation area of the 1050 and 1300 K systems are 641 and 356 sq m respectively. Comparisons are made with Brayton and thermionic systems with both near term and advanced technology considered.

  11. Neutronics analysis of an open-cycle high-impulse gas core reactor concept

    NASA Technical Reports Server (NTRS)

    Whitmarsh, C. L., Jr.

    1972-01-01

    A procedure was developed to calculate the critical fuel mass, including the effects of propellant pressure, for coaxial-flow gas-core reactors operating at 196,600 newtons thrust and 4400 seconds specific impulse. Data were generated for a range of cavity diameter, reflector-moderator thickness, and quantity of structural material. Also presented are such core characteristics as upper limits on cavity pressure, spectral hardening in very-high-temperature hydrogen, and reactivity coefficients.

  12. Preliminary LEU fuel cycle analyses for the Belgian BR2 reactor

    SciTech Connect

    Deen, J.R.; Snelgrove, J.L.

    1986-01-01

    Fuel cycle calculations have been performed with reference HEU fuel and LEU fuel using Cd wires or boron as burnable absorbers. The /sup 235/U content in the LEU element has increased 20% to 480g compared to the reference HEU element. The number of fuel plates has remained unchanged while the fuel meat thickness has increased to 0.76 mm from 0.51 mm. The LEU meat density is 5.1 Mg U/m/sup 3/. The reference fuel cycle was a 31 element core operating at 56 MW with a 19.8 day cycle length and eight fresh elements loaded per cycle. Comparable fuel cycle characteristics can be achieved using the proposed LEU fuel element with either Cd wires or boron burnable absorbers. The neutron flux for E/sub n/ > 1 eV changes very little (<5%) in LEU relative to HEU cores. Thermal flux reductions are 5 to 10% in non-fueled positions, and 20 to 30% in fuel elements.

  13. Heuristic optimization of pressurized water reactor fuel cycle design under general constraints

    SciTech Connect

    Moon, H.; Levine, S.H. ); Mahgerefteh, M. )

    1989-12-01

    Optimization techniques in fuel management have directed modern fuel cycle designs to use low-leakage loading patterns. Future optimization calculations involving low-leakage patterns must utilize nucleonic models that are both fast operationally and rigorous. A two-dimensional two-group diffusion theory code is developed and lattice homogenization constants are generated using a modified LEOPARD code to fulfill these criteria. Based on these two codes, a heuristic optimization study is performed that considers the general constraints (e.g., spent-fuel storage limit and mechanical burnup limit) given to a utility fuel cycle designer. The optimum cycle length that minimizes the fuel cost is {approximately} 600 effective full-power days for the conditions assumed.

  14. Biosurfactant production by Pseudomonas aeruginosa SP4 using sequencing batch reactors: effects of oil loading rate and cycle time.

    PubMed

    Pornsunthorntawee, Orathai; Maksung, Sasiwan; Huayyai, Onsiri; Rujiravanit, Ratana; Chavadej, Sumaeth

    2009-01-01

    In this present study, sequencing batch reactors (SBRs) were used for biosurfactant production from Pseudomonasaeruginosa SP4, which was isolated from petroleum-contaminated soil in Thailand. Two identical lab-scale aerobic SBR units were operated at a constant temperature of 37 degrees C, and a mineral medium (MM) with palm oil was used as the culture medium. The effects of oil loading rate (OLR) and cycle time on the biosurfactant production were studied. The results indicated that the optimum conditions for the biosurfactant production were at an OLR of 2 kg/m(3)days and a cycle time of 2 days/cycle, which provided a surface tension reduction of 59%, a chemical oxygen demand (COD) removal of 90%, and an oil removal of 97%. Under the optimum conditions, it was found that the biosurfactant production was maximized at an aeration time of 40 h. These preliminary results suggest that the SBR can potentially be adapted for biosurfactant production, and perhaps further developed, potentially for large-scale biosurfactant production. PMID:18672362

  15. Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities, Sections 15-19

    SciTech Connect

    Schneider, K.J.

    1982-09-01

    Information is presented under the following section headings: fuel reprocessing; spent fuel and high-level and transuranic waste storage; spent fuel and high-level and transuranic waste disposal; low-level and intermediate-level waste disposal; and, transportation of radioactive materials in the nuclear fuel cycle. In each of the first three sections a description is given on the mainline process, effluent processing and waste management systems, plant layout, and alternative process schemes. Safety information and a summary are also included in each. The section on transport of radioactive materials includes information on the transportation of uranium ore, uranium ore concentrate, UF/sub 6/, PuO/sub 2/ powder, unirradiated uranium and mixed-oxide fuel assemblies, spent fuel, solidified high-level waste, contact-handled transuranic waste, remote-handled transuranic waste, and low and intermediate level nontransuranic waste. A glossary is included. (JGB)

  16. REACTOR

    DOEpatents

    Spitzer, L. Jr.

    1962-01-01

    The system conteraplates ohmically heating a gas to high temperatures such as are useful in thermonuclear reactors of the stellarator class. To this end the gas is ionized and an electric current is applied to the ionized gas ohmically to heat the gas while the ionized gas is confined to a central portion of a reaction chamber. Additionally, means are provided for pumping impurities from the gas and for further heating the gas. (AEC)

  17. Nuclear proliferation and civilian nuclear power. Report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle description

    SciTech Connect

    Not Available

    1980-06-01

    The Nonproliferation Alterntive Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates.

  18. High cycle fatigue behavior of Incoloy 800H in a simulated high-temperature gas-cooled reactor helium environment

    SciTech Connect

    Soo, P.; Sabatini, R.L.; Epel, L.G.; Hare, J.R. Sr.

    1980-01-01

    The current study was an attempt to evaluate the high cycle fatigue strength of Incoloy 800H in a High-Temperature Gas-Cooled Reactor helium environment containing significant quantities of moisture. As-heat-treated and thermally-aged materials were tested to determine the effects of long term corrosion in the helium test gas. Results from in-helium tests were compared to those from a standard air environment. It was found that the mechanisms of fatigue failure were very complex and involved recovery/recrystallization of the surface ground layer on the specimens, sensitization, hardness changes, oxide scale integrity, and oxidation at the tips of propagation cracks. For certain situations a corrosion-fatigue process seems to be controlling. However, for the helium environment studied, there was usually no aging or test condition for which air gave a higher fatigue strength.

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

    NASA Technical Reports Server (NTRS)

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

    1975-01-01

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

  20. Assessment of Possible Cycle Lengths for Fully Encapsulated Microstructure fueled light water reactor Concepts

    SciTech Connect

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal O. Pasamehmetoglu

    2013-02-01

    The use of TRISO-particle-based dispersion fuel within SiC matrix and cladding materials has the potential to allow the design of extremely safe LWRs with failure-proof fuel. This paper examines the feasibility of LWR-like cycle length for such fuel with the imposed constraint of strictly retaining the original geometry of the fuel pins and assemblies. The motivation for retaining the original geometry is to provide the ability to incorporate the fuel “as-is” into existing LWRs while retaining their thermal–hydraulic characteristics. Another mandatory constraint is use of low enriched uranium (at or below 20 w/o). The feasibility of using this fuel is assessed by looking at two factors: cycle lengths and fuel material failure rates. Other considerations (e.g., safety parameters such as reactivity coefficients, feedback, etc.) were not considered at this stage of the study. The study includes the examination of increases in the TRISO kernel sizes without changing the thickness of any of the coating layers. In addition, cases where the buffer layer thickness is allowed to vary are also considered. The study shows that a naïve use of UO2 (even up to 20 w/o enrichment) results in cycle lengths too short to be practical for existing LWR designs and operational demands. Increasing fissile inventory within the fuel compacts shows that acceptable cycle lengths can be achieved. The increase of fissile inventory can be accomplished through multiple means, including higher particle packing fraction, higher enrichment, larger fuel kernel sizes, and the use of higher density fuels (that contain a higher number of U atoms per unit volume). In this study, starting with the recognized highest packing fraction practically achievable (44%), combinations of the other means have been evaluated. The models demonstrate cycle lengths comparable to those of ordinary LWRs. As expected, TRISO particles with extremely large kernels are shown to fail under all considered scenarios. In

  1. A description of the demonstration Integral Fast Reactor fuel cycle facility.

    PubMed

    Courtney, J C; Carnes, M D; Dwight, C C; Forrester, R J

    1991-10-01

    A fuel examination facility at the Idaho National Engineering Laboratory is being converted into a facility that will electrochemically process spent fuel. This is an important step in the demonstration of the Integral Fast Reactor concept being developed by Argonne National Laboratory. Renovations are designed to bring the facility up to current health and safety and environmental standards and to support its new mission. Improvements include the addition of high-reliability earthquake hardened off-gas and electrical power systems, the upgrading of radiological instrumentation, and the incorporation of advances in contamination control. A major task is the construction of a new equipment repair and decontamination facility in the basement of the building to support operations. PMID:1917494

  2. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  3. Contribution of Clinch River Breeder Reactor plant design and development to the LMFBR fuel cycle

    SciTech Connect

    Riley, D.R.; Dickson, P.W.

    1981-01-01

    This paper describes how the CRBRP development and CRBRP focus of the LMFBR base technology program have led to advances in the state of the art in physics, thermal-hydraulics, structural analysis, core restraint, seismic analysis, and analysis of hypothetical core-disruptive accident energetics, all of which have been incorporated through disciplined engineering into the final CRBRP design. The total development in the US of fuels and materials, the analytical advances made on CRBRP design, and the incorporation of the latest experimental results into that design have put the US technology in general and the CRBRP design in particular at the forefront of technology. This has placed the US in a position to develop the most favorable LMFBR fuel cycle.

  4. Mathematical modelling and reactor design for multi-cycle bioregeneration of nitrate exhausted ion exchange resin.

    PubMed

    Ebrahimi, Shelir; Roberts, Deborah J

    2016-01-01

    Nitrate contamination is one of the largest issues facing communities worldwide. One of the most common methods for nitrate removal from water is ion exchange using nitrate selective resin. Although these resins have a great capacity for nitrate removal, they are considered non regenerable. The sustainability of nitrate-contaminated water treatment processes can be achieved by regenerating the exhausted resin several times rather than replacing and incineration of exhausted resin. The use of multi-cycle exhaustion/bioregeneration of resin enclosed in a membrane has been shown to be an effective and innovative regeneration method. In this research, the mechanisms for bioregeneration of resin were studied and a mathematical model which incorporated physical desorption process with biological removal kinetics was developed. Regardless of the salt concentration of the solution, this specific resin is a pore-diffusion controlled process (XδD ¯CDr0(5+2α)<1). Also, Thiele modulus was calculated to be between 4 and 12 depending on the temperature and salt concentration. High Thiele modulus (>3) shows that the bioregeneration process is controlled by reaction kinetics and is governed by biological removal of nitrate. The model was validated by comparison to experimental data; the average of R-squared values for cycle 1 to 5 of regeneration was 0.94 ± 0.06 which shows that the developed model predicted the experimental results very well. The model sensitivity for different parameters was evaluated and a model bioreactor design for bioregeneration of highly selective resins was also presented. PMID:26595098

  5. Low cycle fatigue behavior of a ferritic reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Sarkar, Apu; Kumawat, Bhupendra K.; Chakravartty, J. K.

    2015-07-01

    The cyclic stress-strain response and the low cycle fatigue (LCF) behavior of 20MnMoNi55 pressure vessel steel were studied. Tensile strength and LCF properties were examined at room temperature (RT) using specimens cut from rolling direction of a rolled block. The fully reversed strain-controlled LCF tests were conducted at a constant total strain rate with different axial strain amplitude levels. The cyclic strain-stress relationships and the strain-life relationships were obtained through the test results, and related LCF parameters of the steel were calculated. The studied steel exhibits cyclic softening behavior. Furthermore, analysis of stabilized hysteresis loops showed that the steel exhibits non-Masing behavior. Complementary scanning electron microscopy examinations were also carried out on fracture surfaces to reveal dominant damage mechanisms during crack initiation, propagation and fracture. Multiple crack initiation sites were observed on the fracture surface. The investigated LCF behavior can provide reference for pressure vessel life assessment and fracture mechanisms analysis.

  6. High-Temperature Gas-Cooled Reactor Steam Cycle/Cogeneration Lead Project strategy plan

    SciTech Connect

    1982-03-01

    The strategy for developing the HTGR system and introducing it into the energy marketplace is based on using the most developed technology path to establish a HTGR-Steam Cycle/Cogeneration (SC/C) Lead Project. Given the status of the HTGR-SC/C technology, a Lead Plant could be completed and operational by the mid 1990s. While there is remaining design and technology development that must be accomplished to fulfill technical and licensing requirements for a Lead Project commitment, the major barriers to the realization a HTGR-SC/C Lead Project are institutional in nature, e.g. Project organization and management, vendor/supplier development, cost/risk sharing between the public and private sector, and Project financing. These problems are further exacerbated by the overall pervading issues of economic and regulatory instability that presently confront the utility and nuclear industries. This document addresses the major institutional issues associated with the HTGR-SC/C Lead Project and provides a starting point for discussions between prospective Lead Project participants toward the realization of such a Project.

  7. Renovation of Chemical Processing Facility for Development of Advanced Fast Reactor Fuel Cycle System in JNC

    SciTech Connect

    Atsushi Aoshima; Shigehiko Miyachi; Takashi Suganuma; Shinichi Nemoto

    2002-07-01

    The CPF had 4 laboratories (operation room A, laboratory A, laboratory C and analysis laboratory) in connection with reprocessing technology. The main laboratory, operation room A, has 5 hot cells. Since equipment in the main cell had been designed for small-scale verification of existing reprocessing steps, it was hardly able to respond flexibly to experimental studies on advanced technology. It was decided to remodel the cell according to the design that was newly laid out in order to ensure the function and space to conduct various basic tests. The other laboratories had no glove boxes for conducting basic experiments of important elements in the advanced reprocessing, such as actinides except U and Pu, lanthanides and so on. In order to meet various requirements of innovative technologies on advanced fuel cycle development, one laboratory is established more for study on dry reprocessing, and glove boxes, hoods and analytical equipment such as NMR, FT-IR, TI-MS are newly installed in the other laboratories in this renovation. After the renovation, hot tests in the CPF will be resumed from April 2002. (authors)

  8. Enhancing digestion efficiency of POME in anaerobic sequencing batch reactor with ozonation pretreatment and cycle time reduction.

    PubMed

    Chaiprapat, Sumate; Laklam, Tanyaluk

    2011-03-01

    Ozonation pretreatment was applied to palm oil mill effluent (POME) prior to anaerobic digestion using the anaerobic sequencing batch reactor (ASBR). Ozonation increased BOD/COD by 37.9% with a COD loss of only 3.3%. At organic loads of 6.48-12.96 kg COD/m(3)/d, feeding with non-ozonated POME caused a system failure. The ozonated POME gave significantly higher TCOD removal at loadings 6.52 and 9.04 kg COD/m(3)/d but failed to sustain the operation at loading 11.67 kg COD/m(3)/d. Effects of cycle time (CT) and hydraulic retention time (HRT) were determined using quadratic regression model. The generated response surface and contour plot showed that at this high load conditions (6.52-11.67 kg COD/m(3)/d), longer HRT and shorter CT gave the ASBR higher organic removal efficiency and methane yield. The model was able to satisfactorily describe the relationship of these two key operating parameters. PMID:21215615

  9. Life cycle assessment of hydrogen production from S-I thermochemical process coupled to a high temperature gas reactor

    SciTech Connect

    Giraldi, M. R.; Francois, J. L.; Castro-Uriegas, D.

    2012-07-01

    The purpose of this paper is to quantify the greenhouse gas (GHG) emissions associated to the hydrogen produced by the sulfur-iodine thermochemical process, coupled to a high temperature nuclear reactor, and to compare the results with other life cycle analysis (LCA) studies on hydrogen production technologies, both conventional and emerging. The LCA tool was used to quantify the impacts associated with climate change. The product system was defined by the following steps: (i) extraction and manufacturing of raw materials (upstream flows), (U) external energy supplied to the system, (iii) nuclear power plant, and (iv) hydrogen production plant. Particular attention was focused to those processes where there was limited information from literature about inventory data, as the TRISO fuel manufacture, and the production of iodine. The results show that the electric power, supplied to the hydrogen plant, is a sensitive parameter for GHG emissions. When the nuclear power plant supplied the electrical power, low GHG emissions were obtained. These results improve those reported by conventional hydrogen production methods, such as steam reforming. (authors)

  10. Sensitivities of WRF-Chem to dust emission schemes and land surface properties in simulating dust cycles during springtime over East Asia

    NASA Astrophysics Data System (ADS)

    Su, Lin; Fung, Jimmy C. H.

    2015-11-01

    The sensitivities of Weather Research and Forecasting model coupled with chemistry (WRF-Chem) to the Air Force Weather Agency (AFWA) and Shao2011 (S11) dust emission schemes, and to various land surface properties generated from United States Geological Survey (USGS) and Beijing Normal University (BNU) soil data over East Asia for spring 2012, are examined in this study. The results show that the dust emissions generated with the S11 scheme are 2-5 times that generated with the AFWA emission scheme, with emissions ranging from 0.2-1 Tg/d over East Asia in the relevant period. The AFWA emission scheme omits almost the entire Gobi desert and produces low dust emissions, whereas large amounts of dust emission in this region are produced with the S11 emission scheme, regardless of whether USGS or BNU soil data are used. The surface particulate matter 10 (PM10) concentrations are reasonably well reproduced by the model with different configurations. However, the S11 emission scheme has better performance in simulating the surface PM10 concentrations than the AFWA scheme, especially near the Gobi desert. Comparisons with satellite-based observations suggest that WRF-Chem gives better performance with S11 emission scheme in reproducing the horizontal and vertical distribution of aerosol optical properties. The discrepancy between the performances of the AFWA and S11 emission schemes is mainly due to the underestimation of the dust emission over the Gobi desert by the AFWA scheme, which scales the dust emission directly based on the erodibility factor, indicating that the erodibility factor over the Gobi desert is highly underestimated and highlighting an urgent need to improve the erodibility data set.

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

    SciTech Connect

    Moisseytsev, A.; Sienicki, J. J.

    2007-03-08

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

  12. Progress and interim results of the INPRO joint study on assessment of INS based on closed nuclear fuel cycle with fast reactors

    SciTech Connect

    Usanov, Vladimir; Raj, Baldev; Vasile, Alfredo

    2007-07-01

    The purpose of the work is to review interim results of the Joint Study on assessment of an Innovative Nuclear System based on a Closed Nuclear Fuel Cycle with Fast Reactors (INS CNFC-FR). This study is a part of the IAEA international project for innovative reactors and fuel cycle technologies (INPRO). Now it is being implemented by Canada, China, France, India, Japan, Republic of Korea, Russia, and Ukraine. A report on results of implementation of the first phase of the Joint Study was presented to the INPRO Steering Committee meeting in December 2006. It was also agreed by the Joint Study participants to reveal these results to broader discussion at scientific conferences and meetings. The authors' interpretation of the Joint Study findings and issues is presented in the paper. (authors)

  13. The effectiveness of using the combined-cycle technology in a nuclear power plant unit equipped with an SVBR-100 reactor

    NASA Astrophysics Data System (ADS)

    Kasilov, V. F.; Dudolin, A. A.; Gospodchenkov, I. V.

    2015-05-01

    The design of a modular SVBR-100 reactor with a lead-bismuth alloy liquid-metal coolant is described. The basic thermal circuit of a power unit built around the SVBR-100 reactor is presented together with the results of its calculation. The gross electrical efficiency of the turbine unit driven by saturated steam at a pressure of 6.7 MPa is estimated at η{el/gr} = 35.5%. Ways for improving the efficiency of this power unit and increasing its power output by applying gas-turbine and combined-cycle technologies are considered. With implementing a combined-cycle power-generating system comprising two GE-6101FA gas-turbine units with a total capacity of 140 MW, it becomes possible to obtain the efficiency of the combined-cycle plant equipped with the SVBR-100 reactor η{el/gr} = 45.39% and its electrical power output equal to 328 MW. The heat-recovery boiler used as part of this power installation generates superheated steam with a temperature of 560°C, due to which there is no need to use a moisture separator/steam reheater in the turbine unit thermal circuit.

  14. Sensitivity analysis and optimization of the nuclear fuel cycle

    SciTech Connect

    Passerini, S.; Kazimi, M. S.; Shwageraus, E.

    2012-07-01

    A sensitivity study has been conducted to assess the robustness of the conclusions presented in the MIT Fuel Cycle Study. The Once Through Cycle (OTC) is considered as the base-line case, while advanced technologies with fuel recycling characterize the alternative fuel cycles. The options include limited recycling in LWRs and full recycling in fast reactors and in high conversion LWRs. Fast reactor technologies studied include both oxide and metal fueled reactors. The analysis allowed optimization of the fast reactor conversion ratio with respect to desired fuel cycle performance characteristics. The following parameters were found to significantly affect the performance of recycling technologies and their penetration over time: Capacity Factors of the fuel cycle facilities, Spent Fuel Cooling Time, Thermal Reprocessing Introduction Date, and in core and Out-of-core TRU Inventory Requirements for recycling technology. An optimization scheme of the nuclear fuel cycle is proposed. Optimization criteria and metrics of interest for different stakeholders in the fuel cycle (economics, waste management, environmental impact, etc.) are utilized for two different optimization techniques (linear and stochastic). Preliminary results covering single and multi-variable and single and multi-objective optimization demonstrate the viability of the optimization scheme. (authors)

  15. Effect of temperature cycling on the activity and productivity of immobilized beta-galactosidase in a thermally reversible hydrogel bead reactor.

    PubMed

    Park, T G; Hoffman, A S

    1988-10-01

    The enzyme beta-galactosidase has been immobilized within thermally reversible hydrogel beads that exhibit LCST (lower critical solution temperature) behavior. The hydrogel beads containing the immobilized enzymes swell and expand below the LCST and deswell and shrink above the LCST. This behavior is reversible. The enzyme was physically entrapped in a crosslinked hydrogel of a copolymer of N-isopropylacrylamide (NIPAAm) and acrylamide (AAm), and formed as beads in an inverse suspension polymerization. The beads were placed in a packed bed column reactor which was operated in a continuous, single pass mode, either isothermally at 30 or 35 degrees C, or with temperature cycling between 30 and 35 degrees C. The thermal cycling significantly enhanced overall reactor enzyme activity relative to isothermal operation at either the higher or lower temperature. It is postulated that mass transfer rates within the hydrogel beads are greatly enhanced by the movement of water in and out of the beads during the expansion or collapse of the polymer chain network as temperature is cycled. PMID:3144242

  16. Dynamic Modeling and Control of Nuclear Reactors Coupled to Closed-Loop Brayton Cycle Systems using SIMULINK™

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.; Sanchez, Travis

    2005-02-01

    The operation of space reactors for both in-space and planetary operations will require unprecedented levels of autonomy and control. Development of these autonomous control systems will require dynamic system models, effective control methodologies, and autonomous control logic. This paper briefly describes the results of reactor, power-conversion, and control models that are implemented in SIMULINK™ (Simulink, 2004). SIMULINK™ is a development environment packaged with MatLab™ (MatLab, 2004) that allows the creation of dynamic state flow models. Simulation modules for liquid metal, gas cooled reactors, and electrically heated systems have been developed, as have modules for dynamic power-conversion components such as, ducting, heat exchangers, turbines, compressors, permanent magnet alternators, and load resistors. Various control modules for the reactor and the power-conversion shaft speed have also been developed and simulated. The modules are compiled into libraries and can be easily connected in different ways to explore the operational space of a number of potential reactor, power-conversion system configurations, and control approaches. The modularity and variability of these SIMULINK™ models provides a way to simulate a variety of complete power generation systems. To date, both Liquid Metal Reactors (LMR), Gas Cooled Reactors (GCR), and electric heaters that are coupled to gas-dynamics systems and thermoelectric systems have been simulated and are used to understand the behavior of these systems. Current efforts are focused on improving the fidelity of the existing SIMULINK™ modules, extending them to include isotopic heaters, heat pipes, Stirling engines, and on developing state flow logic to provide intelligent autonomy. The simulation code is called RPC-SIM (Reactor Power and Control-Simulator).

  17. Nuclear fuel cycle costs

    SciTech Connect

    Burch, W.D.; Haire, M.J.; Rainey, R.H.

    1982-02-01

    The costs for the back-end of the nuclear fuel cycle, which were developed as part of the Nonproliferation Alternative Systems Assessment Program (NASAP), are presented. Total fuel cycle costs are given for the pressurized water reactor once-through and fuel recycle systems, and for the liquid-metal fast breeder reactor system. These calculations show that fuel cycle costs are a small part of the total power costs. For breeder reactors, fuel cycle costs are about half that of the present once-through system. The total power cost of the breeder reactor system is greater than that of light-water reactor at today's prices for uranium and enrichment.

  18. Opportunities to reduce consumption of natural uranium in reactor SVBR-75/100 when changing over to the closed fuel cycle

    SciTech Connect

    Toshinsky, G.I.; Komlev, O.G.; Mel'nikov, K.G.; Novikova, N.N.

    2007-07-01

    The design of reactor SVBR-75/100 allows it to operate using different types of fuel and in different fuel cycles without changing its design and deteriorating its safety characteristics. Fuel-at-once refueling adopted in the design (lack of partial refueling) makes it possible to change the core content at each refueling by using the type of fuel that is the most economically effective at the current stage of nuclear power (NP) development. In the nearest future use of mastered oxide uranium fuel and operating in the opened fuel cycle with postponed reprocessing will be the most economically effective. Changeover to the mixed uranium-plutonium fuel and closed nuclear fuel cycle (NFC) will be economically effective in an event of increase of natural uranium costs when the expenditures for construction of the enterprises on reprocessing the spent nuclear fuel (SNF), re-fabrication of new fuel with plutonium and their operating are less than the corresponding costs of natural uranium, its enrichment costs, the costs of manufacturing fresh uranium fuel and long temporary storage of SNF. At this, it is possible to use both MOX fuel with weapon or reactor plutonium and mixed nitride fuel in case its usage is more profitable. As fast reactors (FR) using uranium fuel and operating in the opened NFC consume much more natural uranium in comparison with thermal reactors (TR), and at the expected high paces of NP development the cheap resources of natural uranium will be exhausted prior to the middle of the century that will cause increase in the uranium cost, the period of FRs operating in the opened NFC must be maximally reduced. However, it should be mentioned that it is difficult to forecast reliably the date when because of the increased cost of natural uranium the NP will lose its competitiveness with electric power using fossil fuel. This is conditioned by the fact that the cost of the NPP produced electricity is less sensitive to the cost of natural uranium in

  19. Extension of the supercritical carbon dioxide brayton cycle to low reactor power operation: investigations using the coupled anl plant dynamics code-SAS4A/SASSYS-1 liquid metal reactor code system.

    SciTech Connect

    Moisseytsev, A.; Sienicki, J. J.

    2012-05-10

    Significant progress has been made on the development of a control strategy for the supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle enabling removal of power from an autonomous load following Sodium-Cooled Fast Reactor (SFR) down to decay heat levels such that the S-CO{sub 2} cycle can be used to cool the reactor until decay heat can be removed by the normal shutdown heat removal system or a passive decay heat removal system such as Direct Reactor Auxiliary Cooling System (DRACS) loops with DRACS in-vessel heat exchangers. This capability of the new control strategy eliminates the need for use of a separate shutdown heat removal system which might also use supercritical CO{sub 2}. It has been found that this capability can be achieved by introducing a new control mechanism involving shaft speed control for the common shaft joining the turbine and two compressors following reduction of the load demand from the electrical grid to zero. Following disconnection of the generator from the electrical grid, heat is removed from the intermediate sodium circuit through the sodium-to-CO{sub 2} heat exchanger, the turbine solely drives the two compressors, and heat is rejected from the cycle through the CO{sub 2}-to-water cooler. To investigate the effectiveness of shaft speed control, calculations are carried out using the coupled Plant Dynamics Code-SAS4A/SASSYS-1 code for a linear load reduction transient for a 1000 MWt metallic-fueled SFR with autonomous load following. No deliberate motion of control rods or adjustment of sodium pump speeds is assumed to take place. It is assumed that the S-CO{sub 2} turbomachinery shaft speed linearly decreases from 100 to 20% nominal following reduction of grid load to zero. The reactor power is calculated to autonomously decrease down to 3% nominal providing a lengthy window in time for the switchover to the normal shutdown heat removal system or for a passive decay heat removal system to become effective. However, the

  20. Breeding of {sup 233}U in the thorium–uranium fuel cycle in VVER reactors using heavy water

    SciTech Connect

    Marshalkin, V. E. Povyshev, V. M.

    2015-12-15

    A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the {sup 233}U–{sup 232}Th oxide fuel of water-moderated reactors with variable water composition (D{sub 2}O, H{sub 2}O) that ensures breeding of the {sup 233}U and {sup 235}U isotopes. The method is comparatively simple to implement.

  1. Breeding of 233U in the thorium-uranium fuel cycle in VVER reactors using heavy water

    NASA Astrophysics Data System (ADS)

    Marshalkin, V. E.; Povyshev, V. M.

    2015-12-01

    A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the 233U-232Th oxide fuel of water-moderated reactors with variable water composition (D2O, H2O) that ensures breeding of the 233U and 235U isotopes. The method is comparatively simple to implement.

  2. Influence of the cycle length on the production of PHA and polyglucose from glycerol by bacterial enrichments in sequencing batch reactors.

    PubMed

    Moralejo-Gárate, Helena; Palmeiro-Sánchez, Tania; Kleerebezem, Robbert; Mosquera-Corral, Anuska; Campos, José Luis; van Loosdrecht, Mark C M

    2013-12-01

    PHA, a naturally occurring biopolymer produced by a wide range of microorganisms, is known for its applications as bioplastic. In recent years the use of agro-industrial wastewater as substrate for PHA production by bacterial enrichments has attracted considerable research attention. Crude glycerol as generated during biodiesel production is a waste stream that due to its high organic matter content and low price could be an interesting substrate for PHA production. Previously we have demonstrated that when glycerol is used as substrate in a feast-famine regime, PHA and polyglucose are simultaneously produced as storage polymers. The work described in this paper aimed at understanding the effect of the cycle length on the bacterial enrichment process with emphasis on the distribution of glycerol towards PHA and polyglucose. Two sequencing batch reactors where operated with the same hydraulic and biomass retention time. A short cycle length (6 h) favored polyglucose production over PHA, whereas at long cycle length (24 h) PHA was more favored. In both communities the same microorganism appeared dominating, suggesting a metabolic rather than a microbial competition response. Moreover, the presence of ammonium during polymer accumulation did not influence the maximum amount of PHA that was attained. PMID:23835920

  3. High-temperature, high-pressure testing of zinc titanate in a bench-scale fluidized-bed reactor for 100 cycles

    SciTech Connect

    Gupta, R.P.; Gangwal, S.K.

    1993-06-01

    Integrated gasification combined cycle (IGCC) power plants are being advanced worldwide to produce electricity from coal owing to their potential for superior environmental performance, economics, and efficiency in comparison to conventional coal-based power plants. A key component of these plants is a hot-gas desulfurization system employing efficient regenerable mixed-metal oxide sorbents. Leading sorbent candidates include zinc ferrite and zinc titanate. These sorbents can remove hydrogen sulfide (H{sub 2}S) in the fuel gas down to very low levels (typically <20 ppmv) at 500 to 750{degree}C and can be readily regenerated for multicycle operation with air. To this end, the Research Triangle Institute (RTI) has formulated and tested a series of zinc titanate sorbents in a high-temperature, high- pressure HTHP fluidized-bed bench-scale reactor. Multicycle HTHP bench-scale testing of these sorbents under a variety of conditions culminated in the development of a ZT-4 sorbent that exhibited the best overall performance in terms of chemical reactivity, sulfur capacity, regenerability, structural properties, and attrition resistance. Following this parametric study, a life-cycle test consisting of 100 sulfidation-regeneration cycles was carried out with ZT-4 in the bench unit.

  4. Assessment of possible cycle lengths for fully-ceramic micro-encapsulated fuel-based light water reactor concepts

    SciTech Connect

    Sen, R. S.; Pope, M. A.; Ougouag, A. M.; Pasamehmetoglu, K.; Venneri, F.

    2012-07-01

    The use of TRISO-particle-based dispersion fuel within SiC matrix and cladding materials has the potential to allow the design of extremely safe LWRs with accident-tolerant fuel. This paper examines the feasibility of LWR-like cycle length for such a low enriched uranium fuel with the imposed constraint of strictly retaining the original geometry of the fuel pins and assemblies. The motivation for retaining the original geometry is to provide the ability to incorporate the fuel 'as-is' into existing LWRs while retaining their thermal-hydraulic characteristics. The feasibility of using this fuel is assessed by looking at cycle lengths and fuel failure rates. Other considerations (e.g., safety parameters, etc.) were not considered at this stage of the study. The study includes the examination of different TRISO kernel diameters without changing the coating layer thicknesses. The study shows that a naive use of UO{sub 2} results in cycle lengths too short to be practical for existing LWR designs and operational demands. Increasing fissile inventory within the fuel compacts shows that acceptable cycle lengths can be achieved. In this study, starting with the recognized highest packing fraction practically achievable (44%), higher enrichment, larger fuel kernel sizes, and the use of higher density fuels have been evaluated. The models demonstrate cycle lengths comparable to those of ordinary LWRs. As expected, TRISO particles with extremely large kernels are shown to fail under all considered scenarios. In contrast, the designs that do not depart too drastically from those of the nominal NGNP HTR fuel TRISO particles are shown to perform satisfactorily and display a high rate of survival under all considered scenarios. Finally, it is recognized that relaxing the geometry constraint will result in satisfactory cycle lengths even using UO{sub 2}-loaded TRISO particles-based fuel with enrichment at or below 20 w/o. (authors)

  5. ASSESSMENT OF POSSIBLE CYCLE LENGTHS FOR FULLY-CERAMIC MICRO-ENCAPSULATED FUEL-BASED LIGHT WATER REACTOR CONCEPTS

    SciTech Connect

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal Pasamehmetoglu; Francesco Venneri

    2012-04-01

    The use of TRISO-particle-based dispersion fuel within SiC matrix and cladding materials has the potential to allow the design of extremely safe LWRs with failure-proof fuel. This paper examines the feasibility of LWR-like cycle length for such a low enriched uranium fuel with the imposed constraint of strictly retaining the original geometry of the fuel pins and assemblies. The motivation for retaining the original geometry is to provide the ability to incorporate the fuel 'as-is' into existing LWRs while retaining their thermal-hydraulic characteristics. The feasibility of using this fuel is assessed by looking at cycle lengths and fuel failure rates. Other considerations (e.g., safety parameters, etc.) were not considered at this stage of the study. The study includes the examination of different TRISO kernel diameters without changing the coating layer thicknesses. The study shows that a naive use of UO{sub 2} results in cycle lengths too short to be practical for existing LWR designs and operational demands. Increasing fissile inventory within the fuel compacts shows that acceptable cycle lengths can be achieved. In this study, starting with the recognized highest packing fraction practically achievable (44%), higher enrichment, larger fuel kernel sizes, and the use of higher density fuels have been evaluated. The models demonstrate cycle lengths comparable to those of ordinary LWRs. As expected, TRISO particles with extremely large kernels are shown to fail under all considered scenarios. In contrast, the designs that do not depart too drastically from those of the nominal NGNP HTR fuel TRISO particles are shown to perform satisfactorily and display a high rates of survival under all considered scenarios. Finally, it is recognized that relaxing the geometry constraint will result in satisfactory cycle lengths even using UO{sub 2}-loaded TRISO particles-based fuel with enrichment at or below 20 w/o.

  6. Analysis of Angular V-Cycle Multigrid Formulation for Three-Dimensional Discrete Ordinates Shielding Problems

    SciTech Connect

    Kucukboyaci, Vefa; Haghighat, Alireza

    2001-06-17

    New angular multigrid formulations have been developed, including the Simplified Angular Multigrid (SAM), Nested Iteration (NI), and V-Cycle schemes, which are compatible with the parallel environment and the adaptive differencing strategy of the PENTRAN three-dimensional parallel S{sub N} code. Through use of the Fourier analysis method for an infinite, homogeneous medium, the effectiveness of the V-Cycle scheme was investigated for different problem parameters including scattering ratio, spatial differencing weights, quadrature order, and mesh size. The theoretical analysis revealed that the V-Cycle scheme is effective for a large range of scattering ratios and is insensitive to mesh size. The effectiveness of the new schemes was also investigated for practical shielding applications such as the Kobayashi benchmark problem and the boiling water reactor core shroud problem.

  7. Summary of the radiological assessment of the fuel cycle for a thorium-uranium carbide-fueled fast breeder reactor

    SciTech Connect

    Tennery, V.J.; Bomar, E.S.; Bond, W.D.; Meyer, H.R.; Morse, L.E.; Till, J.E.; Yalcintas, M.G.

    1980-01-01

    A large fraction of the potential fuel for nuclear power reactors employing fissionable materials exists as ores of thorium. In addition, certain characteristics of a fuel system based on breeding of the fissionable isotope {sup 233}U from thorium offer the possibility of a greater resistance to the diversion of fissionable material for the fabrication of nuclear weapons. This report consolidates into a single source the principal content of two previous reports which assess the radiological environmental impact of mining and milling of thorium ore and of the reprocessing and refabrication of spent FBR thorium-uranium carbide fuel.

  8. Conceptual design study FY 1981: synfuels from fusion - using the tandem mirror reactor and a thermochemical cycle to produce hydrogen

    SciTech Connect

    Krikorian, O.H.

    1982-02-09

    This report represents the second year's effort of a scoping and conceptual design study being conducted for the express purpose of evaluating the engineering potential of producing hydrogen by thermochemical cycles using a tandem mirror fusion driver. The hydrogen thus produced may then be used as a feedstock to produce fuels such as methane, methanol, or gasoline. The main objective of this second year's study has been to obtain some approximate cost figures for hydrogen production through a conceptual design study.

  9. Investigation of the Performance of D2O-Cooled High-Conversion Reactors for Fuel Cycle Calculations

    SciTech Connect

    Hikaru Hiruta; Gilles Youinou

    2013-09-01

    Th-U fueled cores are identified by evaluating the sensitivity coefficients of keff, mass balance, and void coefficient. The effect of advanced iron alloy cladding (i.e., FeCrAl) on the performance of Pu conversion in MOX fueled cores is studied instead of using standard stainless-steel cladding. Variations in clad thickness and coolant-to-fuel volume ratio are also exercised. The use of FeCrAl instead of SS as a cladding alloy reduces the required Pu enrichment and improves the Pu conversion rate primarily due to the absence of nickel in the cladding alloy that results in the reduction of the neutron absorption. Also the difference in void coefficients between SS and FeCrAl alloys is nearly 500 pcm over the entire burnup range. The report also shows sensitivity and uncertainty analyses in order to characterize D2O cooled HCPWRs from different aspects. The uncertainties of integral parameters (keff and void coefficient) for selected reactor cores are evaluated at different burnup points in order to find similarities and trends respect to D2O-HCPWR.

  10. Effect of intermittent aeration cycle on nutrient removal and microbial community in a fluidized bed reactor-membrane bioreactor combo system.

    PubMed

    Guadie, Awoke; Xia, Siqing; Zhang, Zhiqiang; Zeleke, Jemaneh; Guo, Wenshan; Ngo, Huu Hao; Hermanowicz, Slawomir W

    2014-03-01

    Effect of intermittent aeration cycle (IAC=15/45-60/60min) on nutrient removal and microbial community structure was investigated using a novel fluidized bed reactor-membrane bioreactor (FBR-MBR) combo system. FBR alone was found more efficient for removing PO4-P (>85%) than NH4-N (<40%) and chemical oxygen demand (COD<35%). However, in the combo system, COD and NH4-N removals were almost complete (>98%). Efficient nitrification, stable mixed liquor suspended solid and reduced transmembrane pressure was also achieved. Quantitative real-time polymerase chain reaction results of total bacteria 16S rRNA gene copies per mL of mixed-liquor varied from (2.48±0.42)×10(9) initial to (2.74±0.10)×10(8), (6.27±0.16)×10(9) and (9.17±1.78)×10(9) for 15/45, 45/15 and 60/60min of IACs, respectively. The results of clone library analysis revealed that Proteobacteria (59%), Firmicutes (12%) and Bacteroidetes (11%) were the dominant bacterial group in all samples. Overall, the combo system performs optimum nutrient removal and host stable microbial communities at 45/15min of IAC. PMID:24508900

  11. Plant heat cycles, vessel internal arrangement, and auxiliary systems. Volume five

    SciTech Connect

    Not Available

    1986-01-01

    This volume covers nuclear power plant heat cycles (type of nuclear power cycles, power cycle refinements, BWR/PWR power cycle, BWR/PWR reactor coolant system), reactor vessel internal arrangement (reactor vessel features, BWR/PWR reactor vessel and internals, BWR/PWR reactor core), reactor auxiliary systems (purpose of reactor auxiliary systems, PWR and BWR reactor auxiliary systems, PWR and BWR control rod drive mechanisms).

  12. Characterization study and five-cycle tests in a fixed-bed reactor of titania-supported nickel oxide as oxygen carriers for the chemical-looping combustion of methane.

    PubMed

    Corbella, Beatriz M; de Diego, Luis F; García-Labiano, Francisco; Adánez, Juan; Palaciost, José M

    2005-08-01

    Recent investigations have shown that in the combustion of carbonaceous compounds CO2 and NOx emissions to the atmosphere can be substantially reduced by using a two stage chemical-looping process. In this process, the reduction stage is undertaken in a first reactor in which the framework oxygen of a reducible inorganic oxide is used, instead of the usual atmospheric oxygen, for the combustion of a carbonaceous compound, for instance, methane. The outlet gas from this reactor is mostly composed of CO2 and steam as reaction products and further separation of these two components can be carried out easily by simple condensation of steam. Then, the oxygen carrier found in a reduced state is transported to a second reactor in which carrier regeneration with air takes place at relatively low temperatures, consequently preventing the formation of thermal NOx. Afterward, the regenerated carrier is carried to the first reactor to reinitiate a new cycle and so on for a number of repetitive cycles, while the carrier is able to withstand the severe chemical and thermal stresses involved in every cycle. In this paper, the performance of titania-supported nickel oxides has been investigated in a fixed-bed reactor as oxygen carriers for chemical-looping combustion of methane. Samples with different nickel oxide contents were prepared by successive incipient wet impregnations, and their performance as oxygen carriers was investigated at 900 degrees C and atmospheric pressure in five-cycle fixed-bed reactor tests using pure methane and pure air for the respective reduction and regeneration stages. The evolution of the outlet gas composition in each stage was followed by gas chromatography, and the involved chemical, structural, and textural changes of the carrier in the reactor bed were studied by using different characterization techniques. From the study, it is deduced that the reactivity of these nickel-based oxygen carriers is in the two involved stages and almost independent

  13. A study of a zone approach to IAEA (International Atomic Energy Agency) safeguards: The low-enriched-uranium zone of a light-water-reactor fuel cycle

    SciTech Connect

    Fishbone, L.G.; Higinbotham, W.A.

    1986-06-01

    At present the IAEA designs its safeguards approach with regard to each type of nuclear facility so that the safeguards activities and effort are essentially the same for a given type and size of nuclear facility wherever it may be located. Conclusions regarding a state are derived by combining the conclusions regarding the effectiveness of safeguards for the individual facilities within a state. In this study it was convenient to define three zones in a state with a closed light-water-reactor nuclear fuel cycle. Each zone contains those facilities or parts thereof which use or process nuclear materials of the same safeguards significance: low-enriched uranium, radioactive spent fuel, or recovered plutonium. The possibility that each zone might be treated as an extended material balance area for safeguards purposes is under investigation. The approach includes defining the relevant features of the facilities in the three zones and listing the safeguards activities which are now practiced. This study has focussed on the fresh-fuel zone, the several facilities of which use or process low-enriched uranium. At one extreme, flows and inventories would be verified at each material balance area. At the other extreme, the flows into and out of the zone and the inventory of the whole zone would be verified. There are a number of possible safeguards approaches which fall between the two extremes. The intention is to develop a rational approach which will make it possible to compare the technical effectiveness and the inspection effort for the facility-oriented approach, for the approach involving the zone as a material balance area, and for some reasonable intermediate safeguards approaches.

  14. Evolution towards Economically Viable Magnetic Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Furth, H. P.

    1996-11-01

    Large pedestrian dinosaurs have long been extinct, while flying dinosaurs have evolved from the archaeopteryx to the common sparrow. Removal of superfluous constraints was the key. In order for soi-disant intelligent life to have emerged on Earth, fusion-power emission from our Sun must have been kept sufficiently feeble and slow-changing (c.f., Bethe's Carbon-Cycle) so as to allow time for non-trivial evolution. By contrast, any economically viable fusion-reactor scheme must use some fast-burning fuel (e.g. D-D,D-T,etc.), so as to elude the economic constraints of excessive single-unit size and cost. The quest for livelier fusion fuel tends to motivate various departures from a strictly thermalized ``Maxwellian'' reactor-plasma distribution. Illustrative material will include specific options for applying the joint resources of the international ``Three-Large-Tokamak Collaboration''.

  15. Nuclear fuel cycle information workshop

    SciTech Connect

    Not Available

    1983-01-01

    This overview of the nuclear fuel cycle is divided into three parts. First, is a brief discussion of the basic principles of how nuclear reactors work; second, is a look at the major types of nuclear reactors being used and world-wide nuclear capacity; and third, is an overview of the nuclear fuel cycle and the present industrial capability in the US.

  16. DESTRUCTIVE EXAMINATION OF 3-CYCLE LWR (LIGHT WATER REACTOR) FUEL RODS FROM TURKEY POINT UNIT 3 FOR THE CLIMAX - SPENT FUEL TEST

    SciTech Connect

    ATKIN SD

    1981-06-01

    The destructive examination results of five light water reactor rods from the Turkey Point Unit 3 reactor are presented. The examinations included fission gas collection and analyses, burnup and hydrogen analyses, and a metallographic evaluation of the fuel, cladding, oxide, and hydrides. The rods exhibited a low fission gas release with all other results appearing representative for pressurized water reator fuel rods with similar burnups (28 GWd/MTU) and operating histories.

  17. Preliminary Study of Burnup Characteristics for a Simplified Small Pebble Bed Reactor

    SciTech Connect

    Irwanto, Dwi; Kato, Yukikata; Obara, Toru; Yamanaka, Ichiro

    2010-06-22

    Simplification of the pebble bed reactor by removing the unloading device from the system was peformed. For this reactor design, a suitable fuel-loading scheme is the Peu a Peu (little by little) fueling scheme. In the Peu a Peu modus, there is no unloading device; as such, the fuels are never discharged and remain at the bottom of the core during reactor operation. This means that the burnup cycle and reactivity is controlled by the addition of fuel. The objectives of the the present study were to find a means of carrying out the exact calculations needed to analyze the Peu a Peu fuel-loading scheme and to optimize the fuel composition, and fuel-loading scheme to achieve better burnup characteristics. The Monte Carlo method is used to perform calculations with high accuracy. Before the calculation of the whole core, the analysis for the infinite geometry was performed. The power generated per mass consumed for each combination of the uranium enrichment and packing fraction was analyzed from the parametric survey. By using the optimal value obtained, a whole-core calculation for the small 20 MWth reactor was performed and the criticality and burnup of this design was analyzed.

  18. Preliminary Study of Burnup Characteristics for a Simplified Small Pebble Bed Reactor

    NASA Astrophysics Data System (ADS)

    Irwanto, Dwi; Kato, Yukikata; Yamanaka, Ichiro; Obara, Toru

    2010-06-01

    Simplification of the pebble bed reactor by removing the unloading device from the system was peformed. For this reactor design, a suitable fuel-loading scheme is the Peu à Peu (little by little) fueling scheme. In the Peu à Peu modus, there is no unloading device; as such, the fuels are never discharged and remain at the bottom of the core during reactor operation. This means that the burnup cycle and reactivity is controlled by the addition of fuel. The objectives of the the present study were to find a means of carrying out the exact calculations needed to analyze the Peu à Peu fuel-loading scheme and to optimize the fuel composition, and fuel-loading scheme to achieve better burnup characteristics. The Monte Carlo method is used to perform calculations with high accuracy. Before the calculation of the whole core, the analysis for the infinite geometry was performed. The power generated per mass consumed for each combination of the uranium enrichment and packing fraction was analyzed from the parametric survey. By using the optimal value obtained, a whole-core calculation for the small 20 MWth reactor was performed and the criticality and burnup of this design was analyzed.

  19. Fast Breeder Reactor studies

    SciTech Connect

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

    1980-07-01

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

  20. Composite centered schemes for multidimensional conservation laws

    SciTech Connect

    Liska, R.; Wendroff, B.

    1998-05-08

    The oscillations of a centered second order finite difference scheme and the excessive diffusion of a first order centered scheme can be overcome by global composition of the two, that is by performing cycles consisting of several time steps of the second order method followed by one step of the diffusive method. The authors show the effectiveness of this approach on some test problems in two and three dimensions.

  1. A classification scheme for LWR fuel assemblies

    SciTech Connect

    Moore, R.S.; Williamson, D.A.; Notz, K.J.

    1988-11-01

    With over 100 light water nuclear reactors operating nationwide, representing designs by four primary vendors, and with reload fuel manufactured by these vendors and additional suppliers, a wide variety of fuel assembly types are in existence. At Oak Ridge National Laboratory, both the Systems Integration Program and the Characteristics Data Base project required a classification scheme for these fuels. This scheme can be applied to other areas and is expected to be of value to many Office of Civilian Radioactive Waste Management programs. To develop the classification scheme, extensive information on the fuel assemblies that have been and are being manufactured by the various nuclear fuel vendors was compiled, reviewed, and evaluated. It was determined that it is possible to characterize assemblies in a systematic manner, using a combination of physical factors. A two-stage scheme was developed consisting of 79 assembly types, which are grouped into 22 assembly classes. The assembly classes are determined by the general design of the reactor cores in which the assemblies are, or were, used. The general BWR and PWR classes are divided differently but both are based on reactor core configuration. 2 refs., 15 tabs.

  2. A sliding wafer-OMVPE scheme for fabricating subnanometer superlattices

    NASA Astrophysics Data System (ADS)

    Minagawa, S.; Satoh, S.; Nakatsuka, S.; Kakibayashi, H.

    1988-03-01

    A sliding wafer-OMVPE (Organometallic Vapor Phase Epitaxy) reactor suitable for growing superlattices is developed. The reactor is a two-channel horizontal reactor with a susceptor placed across the two channels. A slider is used to transport the substrate by sliding it along the susceptor surface from one channel to the other. This scheme makes it possible to set the temperature of the susceptor in each channel independently by utilizing the skin effect of radio wave in graphite. The performance of this scheme is demonstrated by growing a superlattice of ten periods of GaAs(3.5 Å)/AlAs(7 Å).

  3. Practical introduction of thorium fuel cycles

    SciTech Connect

    Kasten, P.R.

    1982-01-01

    The pracitcal introduction of throrium fuel cycles implies that thorium fuel cycles compete economically with uranium fuel cycles in economic nuclear power plants. In this study the reactor types under consideration are light water reactors (LWRs), heavy water reactors (HWRs), high-temperature gas-cooled reactors (HTGRs), and fast breeder reactors (FBRs). On the basis that once-through fuel cycles will be used almost exclusively for the next 20 or 25 years, introduction of economic thorium fuel cycles appears best accomplished by commercial introduction of HTGRs. As the price of natural uranium increases, along with commercialization of fuel recycle, there will be increasing incentive to utilize thorium fuel cycles in heavy water reactors and light water reactors as well as in HTGRs. After FBRs and fuel recycle are commercialized, use of thorium fuel cycles in the blanket of FBRs appears advantageous when fast breeder reactors and thermal reactors operate in a symbiosis mode (i.e., where /sup 233/U bred in the blanket of a fast breeder reactor is utilized as fissile fuel in thermal converter reactors).

  4. Proliferation Resistant Nuclear Reactor Fuel

    SciTech Connect

    Gray, L W; Moody, K J; Bradley, K S; Lorenzana, H E

    2011-02-18

    Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount, and

  5. THEORETICAL BASIS FOR MODELING ELEMENT CYCLING

    EPA Science Inventory

    A biophysical basis for modeling element cycling is described. The scheme consists of element cycles, organisms necessary to completely catalyze all the component reactions, and higher organisms as structurally complex systems and as subsystems of more complex ecosystems, all to ...

  6. Sensitivity Analysis and Optimization of the Nuclear Fuel Cycle: A Systematic Approach

    NASA Astrophysics Data System (ADS)

    Passerini, Stefano

    For decades, nuclear energy development was based on the expectation that recycling of the fissionable materials in the used fuel from today's light water reactors into advanced (fast) reactors would be implemented as soon as technically feasible in order to extend the nuclear fuel resources. More recently, arguments have been made for deployment of fast reactors in order to reduce the amount of higher actinides, hence the longevity of radioactivity, in the materials destined to a geologic repository. The cost of the fast reactors, together with concerns about the proliferation of the technology of extraction of plutonium from used LWR fuel as well as the large investments in construction of reprocessing facilities have been the basis for arguments to defer the introduction of recycling technologies in many countries including the US. In this thesis, the impacts of alternative reactor technologies on the fuel cycle are assessed. Additionally, metrics to characterize the fuel cycles and systematic approaches to using them to optimize the fuel cycle are presented. The fuel cycle options of the 2010 MIT fuel cycle study are re-examined in light of the expected slower rate of growth in nuclear energy today, using the CAFCA (Code for Advanced Fuel Cycle Analysis). The Once Through Cycle (OTC) is considered as the base-line case, while advanced technologies with fuel recycling characterize the alternative fuel cycle options available in the future. The options include limited recycling in L WRs and full recycling in fast reactors and in high conversion LWRs. Fast reactor technologies studied include both oxide and metal fueled reactors. Additional fuel cycle scenarios presented for the first time in this work assume the deployment of innovative recycling reactor technologies such as the Reduced Moderation Boiling Water Reactors and Uranium-235 initiated Fast Reactors. A sensitivity study focused on system and technology parameters of interest has been conducted to test

  7. An assessment of the attractiveness of material associated with thorium/uranium and uranium closed fuel cycles from a safeguards perspective

    SciTech Connect

    Bathke, Charles Gary; Wallace, Richard K; Hase, Kevin R; Sleaford, Brad W; Ebbinghaus, Bartley B; Collins, Brian W; Bradley, Keith S; Prichard, Andrew W; Smith, Brian W

    2010-01-01

    This paper reports the continued evaluation of the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with various proposed nuclear fuel cycles. Specifically, this paper examines two closed fuel cycles. The first fuel cycle examined is a thorium fuel cycle in which a pressurized heavy water reactor (PHWR) is fueled with mixtures of plutonium/thorium and {sup 233}U/thorium. The used fuel is then reprocessed using the THOREX process and the actinides are recycled. The second fuel cycle examined consists of conventional light water reactors (LWR) whose fuel is reprocessed for actinides that are then fed to and recycled until consumed in fast-spectrum reactors: fast reactors and accelerator driven systems (ADS). As reprocessing of LWR fuel has already been examined, this paper will focus on the reprocessing of the scheme's fast-spectrum reactors' fuel. This study will indicate what is required to render these materials as having low utility for use in nuclear weapons. Nevertheless, the results of this paper suggest that all reprocessing products evaluated so far need to be rigorously safeguarded and provided high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE). The methodology and key findings will be presented.

  8. Thermochemical cycles

    NASA Technical Reports Server (NTRS)

    Funk, J. E.; Soliman, M. A.; Carty, R. H.; Conger, W. L.; Cox, K. E.; Lawson, D.

    1975-01-01

    The thermochemical production of hydrogen is described along with the HYDRGN computer program which attempts to rate the various thermochemical cycles. Specific thermochemical cycles discussed include: iron sulfur cycle; iron chloride cycle; and hybrid sulfuric acid cycle.

  9. The problem of optimizing the water chemistry used in the primary coolant circuit of a nuclear power station equipped with VVER reactors under the conditions of longer fuel cycle campaigns and increased capacity of power units

    NASA Astrophysics Data System (ADS)

    Sharafutdinov, R. B.; Kharitonova, N. L.

    2011-05-01

    It is shown that the optimal water chemistry of the primary coolant circuit must be substantiated while introducing measures aimed at increasing the power output in operating power units and for the project called AES-2006/AES TOI (a typical optimized project of a nuclear power station with enhanced information support). The experience gained from operation of PWR reactors with an elongated fuel cycle at an increased level of power is analyzed. Conditions under which boron compounds are locally concentrated on the fuel rod surfaces (the hideout phenomenon) and axial offset anomaly occurs are enlisted, and the influence of lithium on the hideout in the pores of deposits on the surfaces of fuel assemblies is shown.

  10. Status of French reactors

    SciTech Connect

    Ballagny, A.

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  11. The development and application of an improved reactor analysis model for fast reactors

    NASA Astrophysics Data System (ADS)

    Hou, Jia

    Accuracy in neutron cross sections calculation and consistency in reactor physics are fundamental requirements in advanced nuclear reactor design and analysis. The work presented in this dissertation focuses on the development and advanced application of a reactor analysis model with updated cross section libraries that is suitable for online cross section generation for fast reactors. Research has been performed in two areas of interest in reactor physics. The first target of the research is to develop effcient modeling capacity of the 1- D lattice code MICROX-2 for its neutron spectrum calculation based on Collision Probability Method (CPM). Expanded master cross section libraries have been generated based on updated nuclear data and optimized fine-group energy structure to accommodate both thermal and fast reactor spectra as well as to comply with the need for advanced fuel cycle analysis. After verifying the new libraries, the solution methods have been reviewed and updated, including the update of interpolation scheme for resonance self-shielding factors and improvement of spatial self-shielding models for various fuel assembly geometries. The assessment of the updated lattice calculation models has shown that the prediction accuracy of lattice properties represented by the eigenvalue and reaction rate ratios is improved, especially for fast neutron spectrum lattices of which the importance of neutrons in the unresolved energy range is high. The second target of the research is to improve the accuracy of few-group nuclear cross section generation for the reactor core calculation. A 2-D pin-by-pin lattice model has been developed based on embedded CPM within the framework of the Nodal Expansion Method (NEM), which is capable of modeling the heterogeneity of the fuel assembly. Then, an online cross section generation methodology along with discontinuity factors has been developed based on Iterative Diffusion- Diffusion Methodology (IDDM), which can minimize the

  12. Fast reactors and nuclear nonproliferation

    SciTech Connect

    Avrorin, E.N.; Rachkov, V.I.; Chebeskov, A.N.

    2013-07-01

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (authors)

  13. Tabled Execution in Scheme

    SciTech Connect

    Willcock, J J; Lumsdaine, A; Quinlan, D J

    2008-08-19

    Tabled execution is a generalization of memorization developed by the logic programming community. It not only saves results from tabled predicates, but also stores the set of currently active calls to them; tabled execution can thus provide meaningful semantics for programs that seemingly contain infinite recursions with the same arguments. In logic programming, tabled execution is used for many purposes, both for improving the efficiency of programs, and making tasks simpler and more direct to express than with normal logic programs. However, tabled execution is only infrequently applied in mainstream functional languages such as Scheme. We demonstrate an elegant implementation of tabled execution in Scheme, using a mix of continuation-passing style and mutable data. We also show the use of tabled execution in Scheme for a problem in formal language and automata theory, demonstrating that tabled execution can be a valuable tool for Scheme users.

  14. Evaluating Environmental, Health and Safety Impacts from Two Nuclear Fuel Cycles: A Comparative Analysis of Once-Through Uranium Use and Plutonium Recycle in Light Water Reactors

    NASA Astrophysics Data System (ADS)

    Smith, Bethan L.

    The work presented in this dissertation represents a systems-level approach to investigate potential net impacts with respect to human health and the environment associated with transitioning to the MOC for the U.S. In Chapter 2, an updated systems-level conceptual model of the OTC is presented to more accurately portray the OTC as currently implemented in the U.S. The conceptual model is the basis for estimating the worker collective doses at each operational stage, and the first demonstration of a quantitative comparative radiological impact assessment from expected normal operations is presented. In the course of evaluating worker collective dose associated with modern OTC practices, it was found that the relative contributions from the two grouped operations (front-end operations for preparing reactor fuel and reactor operations) were substantially different from historical data and conventional wisdom. As a bookend to Chapter 2, a summary is provided that describes the nature of the differences and factors that led to these differences. Detailed information of the work as part of the published journal article based off of this corollary work is included as an Appendix (C). In Chapter 3, the study of worker collective doses from the phased introduction of reprocessing in the MOC scenario, and is presented similarly to the results in Chapter 2. MOC performance was also estimated by evaluating the radioactive waste generated that can be disposed and managed through known disposal practices in shallow-land burial. Relative to the OTC, MOC performance with respect to worker collective dose was not discernibly different; while the volume of radioactive waste generated decreased. It was found that although the sheer volume of radioactive waste avoided is large, the waste disposition pathway is known for the majority of this waste. The radioactive waste that requires disposal at a licensed off-site facility is examined in closer detail. The verification process for

  15. Low overhead slipless carrier phase estimation scheme.

    PubMed

    Cheng, Haiquan; Li, Yan; Kong, Deming; Zang, Jizhao; Wu, Jian; Lin, Jintong

    2014-08-25

    Two slipless schemes are compared with application to single carrier 30 Gbaud quadrature phase shift keying (QPSK) system. An equivalent linewidth model considering the phase noise induced by both the laser linewidth and fiber nonlinearity is applied in the performance analysis. The simulation results show that it is possible to mitigate cycle slip (CS) using only 0.39% pilot overhead for the proposed blind carrier phase recovery (CPR) + pilot-symbols-aided phase unwrapping (PAPU) scheme within 1 dB signal-to-noise ratio (SNR) penalty limit at the bit error ratio (BER) of 10(-3) with 4 MHz equivalent linewidth. PMID:25321277

  16. Physics challenges for advanced fuel cycle assessment

    SciTech Connect

    Giuseppe Palmiotti; Massimo Salvatores; Gerardo Aliberti

    2014-06-01

    Advanced fuel cycles and associated optimized reactor designs will require substantial improvements in key research area to meet new and more challenging requirements. The present paper reviews challenges and issues in the field of reactor and fuel cycle physics. Typical examples are discussed with, in some cases, original results.

  17. Indirect visual cryptography scheme

    NASA Astrophysics Data System (ADS)

    Yang, Xiubo; Li, Tuo; Shi, Yishi

    2015-10-01

    Visual cryptography (VC), a new cryptographic scheme for image. Here in encryption, image with message is encoded to be N sub-images and any K sub-images can decode the message in a special rules (N>=2, 2<=K<=N). Then any K of the N sub-images are printed on transparency and stacked exactly, the message of original image will be decrypted by human visual system, but any K-1 of them get no information about it. This cryptographic scheme can decode concealed images without any cryptographic computations, and it has high security. But this scheme lacks of hidden because of obvious feature of sub-images. In this paper, we introduce indirect visual cryptography scheme (IVCS), which encodes sub-images to be pure phase images without visible strength based on encoding of visual cryptography. The pure phase image is final ciphertexts. Indirect visual cryptography scheme not only inherits the merits of visual cryptography, but also raises indirection, hidden and security. Meanwhile, the accuracy alignment is not required any more, which leads to the strong anti-interference capacity and robust in this scheme. System of decryption can be integrated highly and operated conveniently, and its process of decryption is dynamic and fast, which all lead to the good potentials in practices.

  18. University Reactor Matching Grants Program

    SciTech Connect

    John Valentine; Farzad Rahnema; Said Abdel-Khalik

    2003-02-14

    During the 2002 Fiscal year, funds from the DOE matching grant program, along with matching funds from the industrial sponsors, have been used to support research in the area of thermal-hydraulics. Both experimental and numerical research projects have been performed. Experimental research focused on two areas: (1) Identification of the root cause mechanism for axial offset anomaly in pressurized water reactors under prototypical reactor conditions, and (2) Fluid dynamic aspects of thin liquid film protection schemes for inertial fusion reactor chambers. Numerical research focused on two areas: (1) Multi-fluid modeling of both two-phase and two-component flows for steam conditioning and mist cooling applications, and (2) Modeling of bounded Rayleigh-Taylor instability with interfacial mass transfer and fluid injection through a porous wall simulating the ''wetted wall'' protection scheme in inertial fusion reactor chambers. Details of activities in these areas are given.

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

    SciTech Connect

    Su'ud, Zaki; Sekimoto, H.

    2010-06-22

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

  1. Behavior of 241Am in fast reactor systems - a safeguards perspective

    SciTech Connect

    Beddingfield, David H; Lafleur, Adrienne M

    2009-01-01

    Advanced fuel-cycle developments around the world currently under development are exploring the possibility of disposing of {sup 241}Am from spent fuel recycle processes by burning this material in fast reactors. For safeguards practitioners, this approach could potentially complicate both fresh- and spent-fuel safeguards measurements. The increased ({alpha},n) production in oxide fuels from the {sup 241}Am increases the uncertainty in coincidence assay of Pu in MOX assemblies and will require additional information to make use of totals-based neutron assay of these assemblies. We have studied the behavior of {sup 241}Am-bearing MOX fuel in the fast reactor system and the effect on neutron and gamma-ray source-terms for safeguards measurements. In this paper, we will present the results of simulations of the behavior of {sup 241}Am in a fast breeder reactor system. Because of the increased use of MOX fuel in thermal reactors and advances in fuel-cycle designs aimed at americium disposal in fast reactors, we have undertaken a brief study of the behavior of americium in these systems to better understand the safeguards impacts of these new approaches. In this paper we will examine the behavior of {sup 241}Am in a variety of nuclear systems to provide insight into the safeguards implications of proposed Am disposition schemes.

  2. Hydrolysis of CuCl{sub 2} in the Cu-Cl thermochemical cycle for hydrogen production : experimental studies using a spray reactor with an ultrasonic atomizer.

    SciTech Connect

    Ferrandon, M. S.; Lewis, M. A.; Alvarez, F.; Shafirovich, E.; Chemical Sciences and Engineering Division; Univ. of Texas at El Paso

    2010-03-01

    The Cu-Cl thermochemical cycle is being developed as a hydrogen production method. Prior proof-of-concept experimental work has shown that the chemistry is viable while preliminary modeling has shown that the efficiency and cost of hydrogen production have the potential to meet DOE's targets. However, the mechanisms of CuCl{sub 2} hydrolysis, an important step in the Cu-Cl cycle, are not fully understood. Although the stoichiometry of the hydrolysis reaction, 2CuCl{sub 2} + H{sub 2}O {leftrightarrow} Cu{sub 2}OCl{sub 2} + 2HCl, indicates a necessary steam-to-CuCl{sub 2} molar ratio of 0.5, a ratio as high as 23 has been typically required to obtain near 100% conversion of the CuCl{sub 2} to the desired products at atmospheric pressure. It is highly desirable to conduct this reaction with less excess steam to improve the process efficiency. Per Le Chatelier's Principle and according to the available equilibrium-based model, the needed amount of steam can be decreased by conducting the hydrolysis reaction at a reduced pressure. In the present work, the experimental setup was modified to allow CuCl{sub 2} hydrolysis in the pressure range of 0.4-1 atm. Chemical and XRD analyses of the product compositions revealed the optimal steam-to-CuCl{sub 2} molar ratio to be 20-23 at 1 atm pressure. The experiments at 0.4 atm and 0.7 atm showed that it is possible to lower the steam-to-CuCl{sub 2} molar ratio to 15, while still obtaining good yields of the desired products. An important effect of running the reaction at reduced pressure is the significant decrease of CuCl concentration in the solid products, which was not predicted by prior modeling. Possible explanations based on kinetics and residence times are suggested.

  3. Integral Fast Reactor concept

    SciTech Connect

    Till, C.E.; Chang, Y.I.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path.

  4. Advances by the Integral Fast Reactor Program

    SciTech Connect

    Lineberry, M.J.; Pedersen, D.R.; Walters, L.C.; Cahalan, J.E.

    1991-01-01

    The advances by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, improved passive safety, and the development of a prototype fuel cycle facility. 14 refs.

  5. Performance improvement of robots using a learning control scheme

    NASA Technical Reports Server (NTRS)

    Krishna, Ramuhalli; Chiang, Pen-Tai; Yang, Jackson C. S.

    1987-01-01

    Many applications of robots require that the same task be repeated a number of times. In such applications, the errors associated with one cycle are also repeated every cycle of the operation. An off-line learning control scheme is used here to modify the command function which would result in smaller errors in the next operation. The learning scheme is based on a knowledge of the errors and error rates associated with each cycle. Necessary conditions for the iterative scheme to converge to zero errors are derived analytically considering a second order servosystem model. Computer simulations show that the errors are reduced at a faster rate if the error rate is included in the iteration scheme. The results also indicate that the scheme may increase the magnitude of errors if the rate information is not included in the iteration scheme. Modification of the command input using a phase and gain adjustment is also proposed to reduce the errors with one attempt. The scheme is then applied to a computer model of a robot system similar to PUMA 560. Improved performance of the robot is shown by considering various cases of trajectory tracing. The scheme can be successfully used to improve the performance of actual robots within the limitations of the repeatability and noise characteristics of the robot.

  6. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

  7. Aperiodicity resulting from two-cycle coupling in the Belousov-Zhabotinskii reaction. III. Analysis of a model of the effect of spatial inhomogeneities at the input ports of a continuous-flow, stirred tank reactor

    NASA Astrophysics Data System (ADS)

    Györgyi, László; Field, Richard J.

    1989-11-01

    Deterministic chaos is a well-established phenomenon in continuous-flow, stirred tank reactor (CSTR) experiments with the oscillatory Belousov-Zhabotinskii (BZ) reaction. However, it has not yet been possible to reproduce the experimentally observed, robust chaos in simulations using realistic models of the homogeneous chemical kinetics of the BZ reaction. That it may be necessary to consider spatial inhomogeneities in modeling the BZ chaos is suggested by the existence of strong stirring effects on the aperiodic behavior and by the difficulty of reproducing chaos under the same conditions in reactors of different physical configuration. Such inhomogeneity might result from a lack of perfect mixing in the CSTR, especially near the inlets, or from diffusion of species at low flow rates from the CSTR reaction mixture into the tips of the inlets. The presence of spatial inhomogeneities allows coupling between essentially independent oscillators, a well-known source of chaos. Such a model using a realistic representation of the BZ kinetics leads to an eight-variable set of ordinary differential equations. Numerical analysis of these equations by continuation methods and by numerical integration shows the existence of broad regions of chaos and various hysteresis effects involving limit cycles, a steady state and/or a strange attractor. Tristability was found in calculations in a narrow flow rate range. The computed behavior appears for parameter values closely related to the values used experimentally to obtain similar phenomena, and the visual similarity of the computed and experimental strange attractors is striking. The experimental routes to chaos, period doubling, intermittency, and secondary Hopf bifurcations are all reproduced in the simulations. The computed and experimental structures of periodic windows observed within chaotic regions also are very similar.

  8. Compact Spreader Schemes

    SciTech Connect

    Placidi, M.; Jung, J. -Y.; Ratti, A.; Sun, C.

    2014-07-25

    This paper describes beam distribution schemes adopting a novel implementation based on low amplitude vertical deflections combined with horizontal ones generated by Lambertson-type septum magnets. This scheme offers substantial compactness in the longitudinal layouts of the beam lines and increased flexibility for beam delivery of multiple beam lines on a shot-to-shot basis. Fast kickers (FK) or transverse electric field RF Deflectors (RFD) provide the low amplitude deflections. Initially proposed at the Stanford Linear Accelerator Center (SLAC) as tools for beam diagnostics and more recently adopted for multiline beam pattern schemes, RFDs offer repetition capabilities and a likely better amplitude reproducibility when compared to FKs, which, in turn, offer more modest financial involvements both in construction and operation. Both solutions represent an ideal approach for the design of compact beam distribution systems resulting in space and cost savings while preserving flexibility and beam quality.

  9. Nonstandard finite difference schemes

    NASA Technical Reports Server (NTRS)

    Mickens, Ronald E.

    1995-01-01

    The major research activities of this proposal center on the construction and analysis of nonstandard finite-difference schemes for ordinary and partial differential equations. In particular, we investigate schemes that either have zero truncation errors (exact schemes) or possess other significant features of importance for numerical integration. Our eventual goal is to bring these methods to bear on problems that arise in the modeling of various physical, engineering, and technological systems. At present, these efforts are extended in the direction of understanding the exact nature of these nonstandard procedures and extending their use to more complicated model equations. Our presentation will give a listing (obtained to date) of the nonstandard rules, their application to a number of linear and nonlinear, ordinary and partial differential equations. In certain cases, numerical results will be presented.

  10. Check-Digit Schemes.

    ERIC Educational Resources Information Center

    Wheeler, Mary L.

    1994-01-01

    Discusses the study of identification codes and check-digit schemes as a way to show students a practical application of mathematics and introduce them to coding theory. Examples include postal service money orders, parcel tracking numbers, ISBN codes, bank identification numbers, and UPC codes. (MKR)

  11. Utilisation of thorium in reactors

    NASA Astrophysics Data System (ADS)

    Anantharaman, K.; Shivakumar, V.; Saha, D.

    2008-12-01

    India's nuclear programme envisages a large-scale utilisation of thorium, as it has limited deposits of uranium but vast deposits of thorium. The large-scale utilisation of thorium requires the adoption of closed fuel cycle. The stable nature of thoria and the radiological issues associated with thoria poses challenges in the adoption of a closed fuel cycle. A thorium fuel based Advanced Heavy Water Reactor (AHWR) is being planned to provide impetus to development of technologies for the closed thorium fuel cycle. Thoria fuel has been loaded in Indian reactors and test irradiations have been carried out with (Th-Pu) MOX fuel. Irradiated thorium assemblies have been reprocessed and the separated 233U fuel has been used for test reactor KAMINI. The paper highlights the Indian experience with the use of thorium and brings out various issues associated with the thorium cycle.

  12. Physical control oriented model of large scale refrigerators to synthesize advanced control schemes. Design, validation, and first control results

    NASA Astrophysics Data System (ADS)

    Bonne, François; Alamir, Mazen; Bonnay, Patrick

    2014-01-01

    In this paper, a physical method to obtain control-oriented dynamical models of large scale cryogenic refrigerators is proposed, in order to synthesize model-based advanced control schemes. These schemes aim to replace classical user experience designed approaches usually based on many independent PI controllers. This is particularly useful in the case where cryoplants are submitted to large pulsed thermal loads, expected to take place in the cryogenic cooling systems of future fusion reactors such as the International Thermonuclear Experimental Reactor (ITER) or the Japan Torus-60 Super Advanced Fusion Experiment (JT-60SA). Advanced control schemes lead to a better perturbation immunity and rejection, to offer a safer utilization of cryoplants. The paper gives details on how basic components used in the field of large scale helium refrigeration (especially those present on the 400W @1.8K helium test facility at CEA-Grenoble) are modeled and assembled to obtain the complete dynamic description of controllable subsystems of the refrigerator (controllable subsystems are namely the Joule-Thompson Cycle, the Brayton Cycle, the Liquid Nitrogen Precooling Unit and the Warm Compression Station). The complete 400W @1.8K (in the 400W @4.4K configuration) helium test facility model is then validated against experimental data and the optimal control of both the Joule-Thompson valve and the turbine valve is proposed, to stabilize the plant under highly variable thermals loads. This work is partially supported through the European Fusion Development Agreement (EFDA) Goal Oriented Training Program, task agreement WP10-GOT-GIRO.

  13. Physical control oriented model of large scale refrigerators to synthesize advanced control schemes. Design, validation, and first control results

    SciTech Connect

    Bonne, François; Bonnay, Patrick

    2014-01-29

    In this paper, a physical method to obtain control-oriented dynamical models of large scale cryogenic refrigerators is proposed, in order to synthesize model-based advanced control schemes. These schemes aim to replace classical user experience designed approaches usually based on many independent PI controllers. This is particularly useful in the case where cryoplants are submitted to large pulsed thermal loads, expected to take place in the cryogenic cooling systems of future fusion reactors such as the International Thermonuclear Experimental Reactor (ITER) or the Japan Torus-60 Super Advanced Fusion Experiment (JT-60SA). Advanced control schemes lead to a better perturbation immunity and rejection, to offer a safer utilization of cryoplants. The paper gives details on how basic components used in the field of large scale helium refrigeration (especially those present on the 400W @1.8K helium test facility at CEA-Grenoble) are modeled and assembled to obtain the complete dynamic description of controllable subsystems of the refrigerator (controllable subsystems are namely the Joule-Thompson Cycle, the Brayton Cycle, the Liquid Nitrogen Precooling Unit and the Warm Compression Station). The complete 400W @1.8K (in the 400W @4.4K configuration) helium test facility model is then validated against experimental data and the optimal control of both the Joule-Thompson valve and the turbine valve is proposed, to stabilize the plant under highly variable thermals loads. This work is partially supported through the European Fusion Development Agreement (EFDA) Goal Oriented Training Program, task agreement WP10-GOT-GIRO.

  14. The role of actinide burning and the Integral Fast Reactor in the future of nuclear power

    SciTech Connect

    Hollaway, W.R.; Lidsky, L.M.; Miller, M.M.

    1990-12-01

    A preliminary assessment is made of the potential role of actinide burning and the Integral Fast Reactor (IFR) in the future of nuclear power. The development of a usable actinide burning strategy could be an important factor in the acceptance and implementation of a next generation of nuclear power. First, the need for nuclear generating capacity is established through the analysis of energy and electricity demand forecasting models which cover the spectrum of bias from anti-nuclear to pro-nuclear. The analyses take into account the issues of global warming and the potential for technological advances in energy efficiency. We conclude, as do many others, that there will almost certainly be a need for substantial nuclear power capacity in the 2000--2030 time frame. We point out also that any reprocessing scheme will open up proliferation-related questions which can only be assessed in very specific contexts. The focus of this report is on the fuel cycle impacts of actinide burning. Scenarios are developed for the deployment of future nuclear generating capacity which exploit the advantages of actinide partitioning and actinide burning. Three alternative reactor designs are utilized in these future scenarios: The Light Water Reactor (LWR); the Modular Gas-Cooled Reactor (MGR); and the Integral Fast Reactor (FR). Each of these alternative reactor designs is described in some detail, with specific emphasis on their spent fuel streams and the back-end of the nuclear fuel cycle. Four separation and partitioning processes are utilized in building the future nuclear power scenarios: Thermal reactor spent fuel preprocessing to reduce the ceramic oxide spent fuel to metallic form, the conventional PUREX process, the TRUEX process, and pyrometallurgical reprocessing.

  15. Particle bed reactor modeling

    NASA Technical Reports Server (NTRS)

    Sapyta, Joe; Reid, Hank; Walton, Lew

    1993-01-01

    The topics are presented in viewgraph form and include the following: particle bed reactor (PBR) core cross section; PBR bleed cycle; fuel and moderator flow paths; PBR modeling requirements; characteristics of PBR and nuclear thermal propulsion (NTP) modeling; challenges for PBR and NTP modeling; thermal hydraulic computer codes; capabilities for PBR/reactor application; thermal/hydralic codes; limitations; physical correlations; comparison of predicted friction factor and experimental data; frit pressure drop testing; cold frit mask factor; decay heat flow rate; startup transient simulation; and philosophy of systems modeling.

  16. Hybridization schemes for clusters

    NASA Astrophysics Data System (ADS)

    Wales, David J.

    The concept of an optimum hybridization scheme for cluster compounds is developed with particular reference to electron counting. The prediction of electron counts for clusters and the interpretation of the bonding is shown to depend critically upon the presumed hybridization pattern of the cluster vertex atoms. This fact has not been properly appreciated in previous work, particularly in applications of Stone's tensor surface harmonic (TSH) theory, but is found to be a useful tool when dealt with directly. A quantitative definition is suggested for the optimum cluster hybridization pattern based directly upon the ease of interpretation of the molecular orbitals, and results are given for a range of species. The relationship of this scheme to the detailed cluster geometry is described using Löwdin's partitioned perturbation theory, and the success and range of application of TSH theory are discussed.

  17. BOILING REACTORS

    DOEpatents

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  18. Beyond Scheme F

    SciTech Connect

    Elliott, C.J.; Fisher, H.; Pepin, J.; Gillmann, R.

    1996-07-01

    Traffic classification techniques were evaluated using data from a 1993 investigation of the traffic flow patterns on I-20 in Georgia. First we improved the data by sifting through the data base, checking against the original video for questionable events and removing and/or repairing questionable events. We used this data base to critique the performance quantitatively of a classification method known as Scheme F. As a context for improving the approach, we show in this paper that scheme F can be represented as a McCullogh-Pitts neural network, oar as an equivalent decomposition of the plane. We found that Scheme F, among other things, severely misrepresents the number of vehicles in Class 3 by labeling them as Class 2. After discussing the basic classification problem in terms of what is measured, and what is the desired prediction goal, we set forth desirable characteristics of the classification scheme and describe a recurrent neural network system that partitions the high dimensional space up into bins for each axle separation. the collection of bin numbers, one for each of the axle separations, specifies a region in the axle space called a hyper-bin. All the vehicles counted that have the same set of in numbers are in the same hyper-bin. The probability of the occurrence of a particular class in that hyper- bin is the relative frequency with which that class occurs in that set of bin numbers. This type of algorithm produces classification results that are much more balanced and uniform with respect to Classes 2 and 3 and Class 10. In particular, the cancellation of errors of classification that occurs is for many applications the ideal classification scenario. The neural network results are presented in the form of a primary classification network and a reclassification network, the performance matrices for which are presented.

  19. Time cycle analysis and simulation of material flow in MOX process layout

    SciTech Connect

    Chakraborty, S.; Saraswat, A.; Danny, K.M.; Somayajulu, P.S.; Kumar, A.

    2013-07-01

    The (U,Pu)O{sub 2} MOX fuel is the driver fuel for the upcoming PFBR (Prototype Fast Breeder Reactor). The fuel has around 30% PuO{sub 2}. The presence of high percentages of reprocessed PuO{sub 2} necessitates the design of optimized fuel fabrication process line which will address both production need as well as meet regulatory norms regarding radiological safety criteria. The powder pellet route has highly unbalanced time cycle. This difficulty can be overcome by optimizing process layout in terms of equipment redundancy and scheduling of input powder batches. Different schemes are tested before implementing in the process line with the help of a software. This software simulates the material movement through the optimized process layout. The different material processing schemes have been devised and validity of the schemes are tested with the software. Schemes in which production batches are meeting at any glove box location are considered invalid. A valid scheme ensures adequate spacing between the production batches and at the same time it meets the production target. This software can be further improved by accurately calculating material movement time through glove box train. One important factor is considering material handling time with automation systems in place.

  20. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1959-10-27

    A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

  1. Research reactors

    SciTech Connect

    Tonneson, L.C.; Fox, G.J.

    1996-04-01

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world`s research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted.

  2. CONVECTION REACTOR

    DOEpatents

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  3. Fragile cycles

    NASA Astrophysics Data System (ADS)

    Bonatti, Ch.; Díaz, L. J.

    We study diffeomorphisms f with heterodimensional cycles, that is, heteroclinic cycles associated to saddles p and q with different indices. Such a cycle is called fragile if there is no diffeomorphism close to f with a robust cycle associated to hyperbolic sets containing the continuations of p and q. We construct a codimension one submanifold of Diff(S×S) that consists of diffeomorphisms with fragile heterodimensional cycles. Our construction holds for any manifold of dimension ⩾4.

  4. Classification Schemes: Developments and Survival.

    ERIC Educational Resources Information Center

    Pocock, Helen

    1997-01-01

    Discusses the growth, survival and future of library classification schemes. Concludes that to survive, a scheme must constantly update its policies, and readily adapt itself to accommodate growing disciplines and changing terminology. (AEF)

  5. One pass core design of a super fast reactor

    SciTech Connect

    Liu, Qingjie; Oka, Yoshiaki

    2013-07-01

    One pass core design for Supercritical-pressure light water-cooled fast reactor (Super FR) is proposed. The whole core is cooled with upward flow in one through flow pattern like PWR. Compared with the previous two pass core design; this new flow pattern can significantly simplify the core concept. Upper core structure, coolant flow scheme as well as refueling procedure are as simple as in PWR. In one pass core design, supercritical-pressure water is at approximately 25.0 MPa and enters the core at 280 C. degrees and is heated up in one through flow pattern upwardly to the average outlet temperature of 500 C. degrees. Great density change in vertical direction can cause significant axial power offset during the cycle. Meanwhile, Pu accumulated in the UO{sub 2} fuel blanket assemblies also introduces great power increase during cycle, which requires large amount of flow for heat removal and makes the outlet temperature of blanket low at the beginning of equilibrium cycle (BOEC). To deal with these issues, some MOX fuel is applied in the bottom region of the blanket assembly. This can help to mitigate the power change in blanket due to Pu accumulation and to increase the outlet temperature of the blanket during cycle. Neutron transport and thermohydraulics coupled calculation shows that this design can satisfy the requirement in the Super FR principle for both 500 C. degrees outlet temperature and negative coolant void reactivity. (authors)

  6. ESCAP mobile training scheme.

    PubMed

    Yasas, F M

    1977-01-01

    In response to a United Nations resolution, the Mobile Training Scheme (MTS) was set up to provide training to the trainers of national cadres engaged in frontline and supervisory tasks in social welfare and rural development. The training is innovative in its being based on an analysis of field realities. The MTS team consisted of a leader, an expert on teaching methods and materials, and an expert on action research and evaluation. The country's trainers from different departments were sent to villages to work for a short period and to report their problems in fulfilling their roles. From these grass roots experiences, they made an analysis of the job, determining what knowledge, attitude and skills it required. Analysis of daily incidents and problems were used to produce indigenous teaching materials drawn from actual field practice. How to consider the problems encountered through government structures for policy making and decisions was also learned. Tasks of the students were to identify the skills needed for role performance by job analysis, daily diaries and project histories; to analyze the particular community by village profiles; to produce indigenous teaching materials; and to practice the role skills by actual role performance. The MTS scheme was tried in Nepal in 1974-75; 3 training programs trained 25 trainers and 51 frontline workers; indigenous teaching materials were created; technical papers written; and consultations were provided. In Afghanistan the scheme was used in 1975-76; 45 participants completed the training; seminars were held; and an ongoing Council was created. It is hoped that the training program will be expanded to other countries. PMID:12265562

  7. Using Steffe's Advanced Fraction Schemes

    ERIC Educational Resources Information Center

    McCloskey, Andrea V.; Norton, Anderson H.

    2009-01-01

    Recognizing schemes, which are different from strategies, can help teachers understand their students' thinking about fractions. Using Steffe's advanced fraction schemes, the authors describe a progression of development that upper elementary and middle school students might follow in understanding fractions. Each scheme can be viewed as a…

  8. Transport reactor development status

    SciTech Connect

    Rush, R.E.; Fankhanel, M.O.; Campbell, W.M.

    1994-10-01

    This project is part of METC`s Power Systems Development Facility (PSDF) located at Wilsonville, Alabama. The primary objective of the Advanced Gasifier module is to produce vitiated gases for intermediate-term testing of Particulate Control Devices (PCDs). The Transport reactor potentially allows particle size distribution, solids loading, and particulate characteristics in the off-gas stream to be varied in a number of ways. Particulates in the hot gases from the Transport reactor will be removed in the PCDs. Two PCDs will be initially installed in the module; one a ceramic candle filter, the other a granular bed filter. After testing of the initial PCDs they will be removed and replaced with PCDs supplied by other vendors. A secondary objective is to verify the performance of a Transport reactor for use in advanced Integrated Gasification Combined Cycle (IGCC), Integrated Gasification Fuel Cell (IG-FC), and Pressurized Combustion Combined Cycle (PCCC) power generation units. This paper discusses the development of the Transport reactor design from bench-scale testing through pilot-scale testing to design of the Process Development Unit (PDU-scale) facility at Wilsonville.

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

    SciTech Connect

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

    2010-07-01

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

  10. Thermal analysis finds optimum FCCU revamp scheme

    SciTech Connect

    Aguilar-Rodriquez, E.; Ortiz-Estrada, C.; Aguilera-Lopez, M. )

    1994-11-07

    The 25,000 b/d fluid catalytic cracking unit (FCCU) at Petroleos Mexicanos' idle Azcapotzalco refinery near Mexico City has been relocated to Pemex's 235,000 b/d Cadereyta refinery. The results of a thermal-integration analysis are being used to revamp the unit and optimize its vapor-recovery scheme. For the case of the Azcapotzalco FCCU, the old unit was designed in the 1950s, so modifications to the reactor/regenerator section incorporate many important changes, including a new riser, feed nozzles, cyclones, air distributor, and other internals. For the new scheme, the analysis was based on the following restrictions: (1) Two cases concerning gas oil feed conditions must be met. In the hot-feed case, feed is introduced from a processing unit outside battery limits (OSBL) at 188 C. For the cold-feed case, feed is introduced from OSBL from storage tanks at 70 C. (2) No new fire heaters are to be installed. (3) Existing equipment must be reused whenever possible. The paper describes and analyzes three alternative schemes.

  11. Menstrual Cycle

    MedlinePlus

    ... Pregnancy This information in Spanish ( en español ) The menstrual cycle Day 1 starts with the first day of ... drop around Day 25 . This signals the next menstrual cycle to begin. The egg will break apart and ...

  12. NEUTRONIC REACTOR

    DOEpatents

    Fraas, A.P.; Mills, C.B.

    1961-11-21

    A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

  13. REACTOR COOLING

    DOEpatents

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  14. Current Comparison of Advanced Nuclear Fuel Cycles

    SciTech Connect

    Steven Piet; Trond Bjornard; Brent Dixon; Robert Hill; Gretchen Matthern; David Shropshire

    2007-04-01

    This paper compares potential nuclear fuel cycle strategies – once-through, recycling in thermal reactors, sustained recycle with a mix of thermal and fast reactors, and sustained recycle with fast reactors. Initiation of recycle starts the draw-down of weapons-usable material and starts accruing improvements for geologic repositories and energy sustainability. It reduces the motivation to search for potential second geologic repository sites. Recycle in thermal-spectru

  15. Comparison of tokamak burn cycle options

    SciTech Connect

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K. Jr.; Hassanein, A.M.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1985-01-01

    Experimental confirmation of noninductive current drive has spawned a number of suggestions as to how this technique can be used to extend the fusion burn period and improve the reactor prospects of tokamaks. Several distinct burn cycles, which employ various combinations of Ohmic and noninductive current generation, are possible, and we will study their relative costs and benefits for both a commerical reactor as well as an INTOR-class device. We begin with a review of the burn cycle options.

  16. Electricity storage using a thermal storage scheme

    SciTech Connect

    White, Alexander

    2015-01-22

    The increasing use of renewable energy technologies for electricity generation, many of which have an unpredictably intermittent nature, will inevitably lead to a greater demand for large-scale electricity storage schemes. For example, the expanding fraction of electricity produced by wind turbines will require either backup or storage capacity to cover extended periods of wind lull. This paper describes a recently proposed storage scheme, referred to here as Pumped Thermal Storage (PTS), and which is based on “sensible heat” storage in large thermal reservoirs. During the charging phase, the system effectively operates as a high temperature-ratio heat pump, extracting heat from a cold reservoir and delivering heat to a hot one. In the discharge phase the processes are reversed and it operates as a heat engine. The round-trip efficiency is limited only by process irreversibilities (as opposed to Second Law limitations on the coefficient of performance and the thermal efficiency of the heat pump and heat engine respectively). PTS is currently being developed in both France and England. In both cases, the schemes operate on the Joule-Brayton (gas turbine) cycle, using argon as the working fluid. However, the French scheme proposes the use of turbomachinery for compression and expansion, whereas for that being developed in England reciprocating devices are proposed. The current paper focuses on the impact of the various process irreversibilities on the thermodynamic round-trip efficiency of the scheme. Consideration is given to compression and expansion losses and pressure losses (in pipe-work, valves and thermal reservoirs); heat transfer related irreversibility in the thermal reservoirs is discussed but not included in the analysis. Results are presented demonstrating how the various loss parameters and operating conditions influence the overall performance.

  17. Electricity storage using a thermal storage scheme

    NASA Astrophysics Data System (ADS)

    White, Alexander

    2015-01-01

    The increasing use of renewable energy technologies for electricity generation, many of which have an unpredictably intermittent nature, will inevitably lead to a greater demand for large-scale electricity storage schemes. For example, the expanding fraction of electricity produced by wind turbines will require either backup or storage capacity to cover extended periods of wind lull. This paper describes a recently proposed storage scheme, referred to here as Pumped Thermal Storage (PTS), and which is based on "sensible heat" storage in large thermal reservoirs. During the charging phase, the system effectively operates as a high temperature-ratio heat pump, extracting heat from a cold reservoir and delivering heat to a hot one. In the discharge phase the processes are reversed and it operates as a heat engine. The round-trip efficiency is limited only by process irreversibilities (as opposed to Second Law limitations on the coefficient of performance and the thermal efficiency of the heat pump and heat engine respectively). PTS is currently being developed in both France and England. In both cases, the schemes operate on the Joule-Brayton (gas turbine) cycle, using argon as the working fluid. However, the French scheme proposes the use of turbomachinery for compression and expansion, whereas for that being developed in England reciprocating devices are proposed. The current paper focuses on the impact of the various process irreversibilities on the thermodynamic round-trip efficiency of the scheme. Consideration is given to compression and expansion losses and pressure losses (in pipe-work, valves and thermal reservoirs); heat transfer related irreversibility in the thermal reservoirs is discussed but not included in the analysis. Results are presented demonstrating how the various loss parameters and operating conditions influence the overall performance.

  18. Steady-state Analysis Model for Advanced Fuelcycle Schemes

    Energy Science and Technology Software Center (ESTSC)

    2006-05-12

    The model was developed as a part of the study, "Advanced Fuel Cycles and Waste Management", which was performed during 2003—2005 by an ad-hoc expert group under the Nuclear Development Committee in the OECD/NEA. The model was designed for an efficient conduct of nuclear fuel cycle scheme cost analyses. It is simple, transparent and offers users the capability to track down the cost analysis results. All the fuel cycle schemes considered in the model aremore » represented in a graphic format and all values related to a fuel cycle step are shown in the graphic interface, i.e., there are no hidden values embedded in the calculations. All data on the fuel cycle schemes considered in the study including mass flows, waste generation, cost data, and other data such as activities, decay heat and neutron sources of spent fuel and high—level waste along time are included in the model and can be displayed. The user can modify easily the values of mass flows and/or cost parameters and see the corresponding changes in the results. The model calculates: front—end fuel cycle mass flows such as requirements of enrichment and conversion services and natural uranium; mass of waste based on the waste generation parameters and the mass flow; and all costs. It performs Monte Carlo simulations with changing the values of all unit costs within their respective ranges (from lower to upper bounds).« less

  19. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1958-04-22

    A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

  20. NEUTRONIC REACTOR

    DOEpatents

    Metcalf, H.E.; Johnson, H.W.

    1961-04-01

    BS>A nuclear reactor incorporating fuel rods passing through a moderator and including tubes of a material of higher Thermal conductivity than the fuel in contact with the fuel is described. The tubes extend beyond the active portion of the reactor into contant with a fiuld coolant.

  1. PID Control Effectiveness for Surface Reactor Concepts

    SciTech Connect

    Dixon, David D.; Marsh, Christopher L.; Poston, David I.

    2007-01-30

    Control of space and surface fission reactors should be kept as simple as possible, because of the need for high reliability and the difficulty to diagnose and adapt to control system failures. Fortunately, compact, fast-spectrum, externally controlled reactors are very simple in operation. In fact, for some applications it may be possible to design low-power surface reactors without the need for any reactor control after startup; however, a simple proportional, integral, derivative (PID) controller can allow a higher performance concept and add more flexibility to system operation. This paper investigates the effectiveness of a PID control scheme for several anticipated transients that a surface reactor might experience. To perform these analyses, the surface reactor transient code FRINK was modified to simulate control drum movements based on bulk coolant temperature.

  2. Design Studies for a Multiple Application Thermal Reactor for Irradiation Experiments (MATRIX)

    SciTech Connect

    Pope, Michael A.; Gougar, Hans D.; Ryskamp, J. M.

    2015-03-01

    The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Should unforeseen circumstances lead to the decommissioning of ATR, the U.S. Government would be left without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. A survey was conducted in order to catalogue the anticipated needs of potential customers. Then, concepts were evaluated to fill the role for this reactor, dubbed the Multi-Application Thermal Reactor Irradiation eXperiments (MATRIX). The baseline MATRIX design is expected to be capable of longer cycle lengths than ATR given a particular batch scheme. The volume of test space in In-Pile-Tubes (IPTs) is larger in MATRIX than in ATR with comparable magnitude of neutron flux. Furthermore, MATRIX has more locations of greater volume having high fast neutron flux than ATR. From the analyses performed in this work, it appears that the lead MATRIX design can be designed to meet the anticipated needs of the ATR replacement reactor. However, this design is quite immature, and therefore any requirements currently met must be re-evaluated as the design is developed further.

  3. An alternative stratigraphic scheme for the Sarawak Basin

    NASA Astrophysics Data System (ADS)

    Mat-Zin, I. C.; Tucker, M. E.

    1999-04-01

    The present stratigraphic scheme for the Tertiary of Sarawak (Ho, K.F., 1978. Stratigraphic framework for oil exploration in Sarawak. Bulletin of the Geological Society of Malaysia, 10, 1-14), which subdivides the entire sedimentary succession into eight sedimentary cycles, is based on the genetic sedimentary cycle concept. Each cycle starts with a transgressive basal part, followed by a regressive unit, which is in turn overlain by the basal transgressive unit of the next cycle. The limitation of the present scheme is, however, in its general applicability; for instance, when one tries to identify the equivalents of marine surfaces within non-marine sediments. This is critical for some areas in Sarawak where the sediments were deposited predominantly within lower coastal plain to upper coastal plain environments, normally barren of foraminifera. Among other inadequacies of the scheme is the lack of basin-wide association between the cycle boundaries and seismic reflectors. In practice cycle boundaries, picked on the composite well logs, often do not agree with the seismic pick. The scheme is, however, well accepted and continues to be used. The alternative stratigraphic scheme for the Sarawak Basin, which will be discussed in this paper, is based on sequence stratigraphic concepts, with the utilisation of unconformity or its correlative conformity as the stratigraphic boundary. This scheme has been generated as the result of a programme of basin-wide seismic mapping of the Sarawak Basin. Seven regional unconformities within the Tertiary sedimentary sequences have been mapped. The sedimentary units between the unconformities can be recognised as 'Sequences'. The oldest unconformity is that between the basement (Belaga Formation) and the overlying Tertiary sediments, mainly of Late Oligocene age. Since the sediments in the basin are mainly of Tertiary age, the oldest unit of the succession is referred to as the Tertiary One Sequence (T1 S). The next younger sequence

  4. A novel sorbent for transport reactors and fluidized bed reactors

    SciTech Connect

    Copeland, R.; Cesario, M.; Gershanovich, Y.; Sibold, J.; Windecker, B.

    1998-12-31

    Coal Fired Gasifier Combined Cycles (GCC) have both high efficiency and very low emissions. GCCs critically need a method of removing the H{sub 2}S produced from the sulfur in the coal from the hot gases. There has been extensive research on hot gas cleanup systems, focused on the use of a zinc oxide based sorbent (e.g., zinc titanate). TDA Research, Inc. (TDA) is developing a novel sorbent with improved attrition resistance for transport reactors and fluidized bed reactors. The authors are testing sorbents at conditions simulating the operating conditions of the Pinon Pine clean coal technology plant. TDA sulfided several different formulations at 538 C and found several that have high sulfur capacity when tested in a fluidized bed reactor. TDA initiated sorbent regeneration at 538 C. The sorbents retained chemical activity with multiple cycles. Additional tests will be conducted to evaluate the best sorbent formulation.

  5. Heterogeneous Transmutation Sodium Fast Reactor

    SciTech Connect

    S. E. Bays

    2007-09-01

    The threshold-fission (fertile) nature of Am-241 is used to destroy this minor actinide by capitalizing upon neutron capture instead of fission within a sodium fast reactor. This neutron-capture and its subsequent decay chain leads to the breeding of even neutron number plutonium isotopes. A slightly moderated target design is proposed for breeding plutonium in an axial blanket located above the active “fast reactor” driver fuel region. A parametric study on the core height and fuel pin diameter-to-pitch ratio is used to explore the reactor and fuel cycle aspects of this design. This study resulted in both non-flattened and flattened core geometries. Both of these designs demonstrated a high capacity for removing americium from the fuel cycle. A reactivity coefficient analysis revealed that this heterogeneous design will have comparable safety aspects to a homogeneous reactor of comparable size. A mass balance analysis revealed that the heterogeneous design may reduce the number of fast reactors needed to close the current once-through light water reactor fuel cycle.

  6. Advances in Tandem Mirror fusion power reactors

    SciTech Connect

    Perkins, L.J.; Logan, B.G.

    1986-05-20

    The Tandem Mirror exhibits several distinctive features which make the reactor embodiment of the principle very attractive: Simple low-technology linear central cell; steady-state operation; high-..beta.. operation; no driven current or disruptions; divertorless operation; direction conversion of end-loss power; low-surface heat loads; and advanced fusion fuel capability. In this paper, we examine these features in connection with two tandem mirror reactor designs, MARS and MINIMARS, and several advanced reactor concepts including the wall-stabilized reactor and the field-reversed mirror. With a novel compact end plug scheme employing octopole stabilization, MINIMARS is expressly designed for short construction times, factory-built modules, and a small (600 MWe) but economic reactor size. We have also configured the design for low radioactive afterheat and inherent/passive safety under LOCA/LOFA conditions, thereby obviating the need for expensive engineered safety systems. In contrast to the complex and expensive double-quadrupole end-cell of the MARS reactor, the compact octopole end-cell of MINIMARS enables ignition to be achieved with much shorter central cell lengths and considerably improves the economy of scale for small (approx.250 to 600 MWe) tandem mirror reactors. Finally, we examine the prospects for realizing the ultimate potential of the tandem mirror with regard to both innovative configurations and novel neutron energy conversion schemes, and stress that advanced fuel applications could exploit its unique reactor features.

  7. Compact Reactor

    SciTech Connect

    Williams, Pharis E.

    2007-01-30

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  8. Plotting and Scheming

    NASA Technical Reports Server (NTRS)

    2004-01-01

    [figure removed for brevity, see original site] Figure 1 [figure removed for brevity, see original site] Figure 2 Click for larger view

    These two graphics are planning tools used by Mars Exploration Rover engineers to plot and scheme the perfect location to place the rock abrasion tool on the rock collection dubbed 'El Capitan' near Opportunity's landing site. 'El Capitan' is located within a larger outcrop nicknamed 'Opportunity Ledge.'

    The rover visualization team from NASA Ames Research Center, Moffett Field, Calif., initiated the graphics by putting two panoramic camera images of the 'El Capitan' area into their three-dimensional model. The rock abrasion tool team from Honeybee Robotics then used the visualization tool to help target and orient their instrument on the safest and most scientifically interesting locations. The blue circle represents one of two current targets of interest, chosen because of its size, lack of dust, and most of all its distinct and intriguing geologic features. To see the second target location, see the image titled 'Plotting and Scheming.'

    The rock abrasion tool is sensitive to the shape and texture of a rock, and must safely sit within the 'footprint' indicated by the blue circles. The rock area must be large enough to fit the contact sensor and grounding mechanism within the area of the outer blue circle, and the rock must be smooth enough to get an even grind within the abrasion area of the inner blue circle. If the rock abrasion tool were not grounded by its support mechanism or if the surface were uneven, it could 'run away' from its target. The rock abrasion tool is location on the rover's instrument deployment device, or arm.

    Over the next few martian days, or sols, the rover team will use these and newer, similar graphics created with more recent, higher-resolution panoramic camera images and super-spectral data from the miniature thermal emission spectrometer. These data will be used to pick the best

  9. NUCLEAR REACTOR

    DOEpatents

    Moore, R.V.; Bowen, J.H.; Dent, K.H.

    1958-12-01

    A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

  10. Prometheus Reactor I&C Software Development Methodology, for Action

    SciTech Connect

    T. Hamilton

    2005-07-30

    The purpose of this letter is to submit the Reactor Instrumentation and Control (I&C) software life cycle, development methodology, and programming language selections and rationale for project Prometheus to NR for approval. This letter also provides the draft Reactor I&C Software Development Process Manual and Reactor Module Software Development Plan to NR for information.

  11. Core Optimization of a Deep-Burn Pebble Bed Reactor

    SciTech Connect

    Brian Boer; Abderrafi M. Ougouag

    2010-06-01

    Achieving a high fuel burnup in the Deep-Burn (DB) pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum as compared to a ’standard’ UO2 fueled core. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. The DB concept focuses on the destruction of spent fuel transuranics in TRISO coated particle fueled gas-cooled reactors with the aim of a fractional fuel burnup of 60-70% in fissions per initial metal atom (FIMA), using a single-pass, multi in-core fuel (re)cycling scheme. In principle, the DB pebble bed concept employs the same reactor designs as the present low enriched uranium core designs, i.e. the 400 MWth Pebble Bed Modular Reactor (PBMR-400). A Pu and Minor Actinide fueled PBMR-400 design serves as the starting point for a core optimization study. The fuel temperature, power peak, temperature reactivity coefficients, and burnup capabilities of the modified designs are analyzed with the PEBBED code. A code-to-code coupling with the PASTA code allows for the analysis of the TRISO fuel performance for both normal and Loss Of Forced Cooling conditions. An improved core design is sought, maximizing the fuel discharge burnup, while retaining negative temperature reactivity feedback coefficients for the entire temperature range and avoiding high fuel temperatures (fuel failure probabilities).

  12. Hampshire Probation Sports Counselling Scheme.

    ERIC Educational Resources Information Center

    Waldman, Keith

    A sports counseling scheme for young people on criminal probation in Hampshire (England) was developed in the 1980s as a partnership between the Sports Council and the Probation Service. The scheme aims to encourage offenders, aged 14 and up, to make constructive use of their leisure time; to allow participants the opportunity to have positive…

  13. Applications of plasma core reactors to terrestrial energy systems

    NASA Technical Reports Server (NTRS)

    Latham, T. S.; Biancardi, F. R.; Rodgers, R. J.

    1974-01-01

    Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times.-

  14. Hybrid Reactor Simulation of Boiling Water Reactor Power Oscillations

    SciTech Connect

    Huang Zhengyu; Edwards, Robert M.

    2003-08-15

    Hybrid reactor simulation (HRS) of boiling water reactor (BWR) instabilities, including in-phase and out-of-phase (OOP) oscillations, has been implemented on The Pennsylvania State University TRIGA reactor. The TRIGA reactor's power response is used to simulate reactor neutron dynamics for in-phase oscillation or the fundamental mode of the reactor modal kinetics for OOP oscillations. The reactor power signal drives a real-time boiling channel simulation, and the calculated reactivity feedback is in turn fed into the TRIGA reactor via an experimental changeable reactivity device. The thermal-hydraulic dynamics, together with first harmonic mode power dynamics, is digitally simulated in the real-time environment. The real-time digital simulation of boiling channel thermal hydraulics is performed by solving constitutive equations for different regions in the channel and is realized by a high-performance personal computer. The nonlinearity of the thermal-hydraulic model ensures the capability to simulate the oscillation phenomena, limit cycle and OOP oscillation, in BWR nuclear power plants. By adjusting reactivity feedback gains for both modes, various oscillation combinations can be realized in the experiment. The dynamics of axially lumped power distribution over the core is displayed in three-dimensional graphs. The HRS reactor power response mimics the BWR core-wide power stability phenomena. In the OOP oscillation HRS, the combination of reactor response and the simulated first harmonic power using shaping functions mimics BWR regional power oscillations. With this HRS testbed, a monitoring and/or control system designed for BWR power oscillations can be experimentally tested and verified.

  15. Irradiation Facilities at the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2005-12-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC – formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world’s data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities1. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens.

  16. NEUTRONIC REACTOR

    DOEpatents

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  17. NEUTRONIC REACTOR

    DOEpatents

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  18. NUCLEAR REACTOR

    DOEpatents

    Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

    1962-10-23

    A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

  19. REACTOR SHIELD

    DOEpatents

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  20. NUCLEAR REACTOR

    DOEpatents

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  1. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.

    1960-04-01

    A nuclear reactor is described consisting of blocks of graphite arranged in layers, natural uranium bodies disposed in holes in alternate layers of graphite blocks, and coolant tubes disposed in the layers of graphite blocks which do not contain uranium.

  2. NUCLEAR REACTOR

    DOEpatents

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  3. NUCLEAR REACTOR

    DOEpatents

    Breden, C.R.; Dietrich, J.R.

    1961-06-20

    A water-soluble non-volatile poison may be introduced into a reactor to nullify excess reactivity. The poison is removed by passing a side stream of the water containing the soluble poison to an evaporation chamber. The vapor phase is returned to the reactor to decrease the concentration of soluble poison and the liquid phase is returned to increase the concentration of soluble poison.

  4. Heterogeneous Recycling in Fast Reactors

    SciTech Connect

    Forget, Benoit; Pope, Michael; Piet, Steven J.; Driscoll, Michael

    2012-07-30

    Current sodium fast reactor (SFR) designs have avoided the use of depleted uranium blankets over concerns of creating weapons grade plutonium. While reducing proliferation risks, this restrains the reactor design space considerably. This project will analyze various blanket and transmutation target configurations that could broaden the design space while still addressing the non-proliferation issues. The blanket designs will be assessed based on the transmutation efficiency of key minor actinide (MA) isotopes and also on mitigation of associated proliferation risks. This study will also evaluate SFR core performance under different scenarios in which depleted uranium blankets are modified to include minor actinides with or without moderators (e.g. BeO, MgO, B4C, and hydrides). This will be done in an effort to increase the sustainability of the reactor and increase its power density while still offering a proliferation resistant design with the capability of burning MA waste produced from light water reactors (LWRs). Researchers will also analyze the use of recycled (as opposed to depleted) uranium in the blankets. The various designs will compare MA transmutation efficiency, plutonium breeding characteristics, proliferation risk, shutdown margins and reactivity coefficients with a current reference sodium fast reactor design employing homogeneous recycling. The team will also evaluate the out-of-core accumulation and/or burn-down rates of MAs and plutonium isotopes on a cycle-by-cycle basis. This cycle-by-cycle information will be produced in a format readily usable by the fuel cycle systems analysis code, VISION, for assessment of the sustainability of the deployment scenarios.

  5. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  6. Cycle Analysis

    Energy Science and Technology Software Center (ESTSC)

    2012-03-20

    1. The Cycle Analysis code is an Microsoft Excel code that performs many different types of thermodynamic cycle analysis for power producing systems. The code will calculate the temperature and pressure and all other thermodynamic properties at the inlet and outlet of each component. The code also calculates the power that is produced, the efficiency, and the heat transported in the heater, gas chiller and recuperators. The code provides a schematic of the loop andmore » provides the temperature and pressure at each location in the loop. The code also provides a T-S (temperature-entropy) diagram of the loop and often it provides an pressure enthalpy plot as well. 2. This version of the code concentrates on supercritical CO2 power cycles, but by simply changing the name of the working fluid many other types of fluids can be analyzed. The Cycle Analysis code provided here contains 18 different types of power cycles. Each cycle is contained in one worksheet or tab that the user can select. The user can change the yellow highlighted regions to perform different thermodynamic cycle analysis.« less

  7. Multigrid method for the equilibrium equations of elasticity using a compact scheme

    NASA Technical Reports Server (NTRS)

    Taasan, S.

    1986-01-01

    A compact difference scheme is derived for treating the equilibrium equations of elasticity. The scheme is inconsistent and unstable. A multigrid method which takes into account these properties is described. The solution of the discrete equations, up to the level of discretization errors, is obtained by this method in just two multigrid cycles.

  8. Solar Thermal Reactor Materials Characterization

    SciTech Connect

    Lichty, P. R.; Scott, A. M.; Perkins, C. M.; Bingham, C.; Weimer, A. W.

    2008-03-01

    Current research into hydrogen production through high temperature metal oxide water splitting cycles has created a need for robust high temperature materials. Such cycles are further enhanced by the use of concentrated solar energy as a power source. However, samples subjected to concentrated solar radiation exhibited lifetimes much shorter than expected. Characterization of the power and flux distributions representative of the High Flux Solar Furnace(HFSF) at the National Renewable Energy Laboratory(NREL) were compared to ray trace modeling of the facility. In addition, samples of candidate reactor materials were thermally cycled at the HFSF and tensile failure testing was performed to quantify material degradation. Thermal cycling tests have been completed on super alloy Haynes 214 samples and results indicate that maximum temperature plays a significant role in reduction of strength. The number of cycles was too small to establish long term failure trends for this material due to the high ductility of the material.

  9. On-Line Schemes For Computing Rotation Angles For SVDS

    NASA Astrophysics Data System (ADS)

    Ercegovac, Milos D.; Lang, Tomas

    1988-01-01

    Two floating-point radix-2 schemes using on-line arithmetic for implementing the direct two-angle method for SVDs are presented. The first scheme is an on-line variant of the cosine/sine approach and is the fastest of the schemes considered: it performs the 2x2 SVD step in about 2n clock cycles. However, it requires a relatively large number of modules; this number is reduced when some modules are reused, resulting in a time of 3n clock cycles. The number of modules of this on-line version is still larger than that of the conventional one, but this is compensated by the smaller number of bit-slices per module and by the digit-serial communication among modules. The corresponding speed-up ratios are of 5 and 3 with respect to a conventional arithmetic implementation. The second scheme uses an on-line CORDIC approach and performs the 2x2 SVD in about 7n clock cycles and is advantageous because it is more time-area efficient. It results in a speed-up of about 2.5 with respect to the conventional CORDIC implementation.

  10. Research reactors - an overview

    SciTech Connect

    West, C.D.

    1997-03-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

  11. NEUTRONIC REACTOR

    DOEpatents

    Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

    1958-09-01

    This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

  12. Inherent safety of minimum-burnup breed and burn reactors

    SciTech Connect

    Qvist, S.; Reenspan, E.

    2012-07-01

    Reactors that aim to sustain the breed and burn (B and B) mode of operation at minimum discharge burnup require excellent neutron economy, Minimum-burnup B and B cores are generally large and feature low neutron leakage probability and a hard neutron spectrum. While highly promising fuel cycles can be achieved with such designs, the very same features are pushing the limits of the core's ability to passively respond safely to unprotected accidents. Low leakage minimum-burnup sodium-cooled B and B cores have a large positive coolant void-worth and coolant temperature reactivity coefficient. In this study, the applicability of major approaches for fast reactor void-worth reduction is evaluated specifically for B and B cores. The design, shuffling scheme and performance of a new metallic-fueled, sodium-cooled minimum burnup B and B core, used as basis for the void-worth reduction analysis, is presented. The analysis shows that reactivity control systems based on passive {sup 6}Li injection during temperature excursions are the only option able to provide negative void-worth without significantly increasing the minimum burnup required for sustaining the B and B mode of operation. A new type of lithium expansion module (LEM) system was developed specifically for B and B cores and its effect on core performance is presented. (authors)

  13. Cycling injuries.

    PubMed Central

    Cohen, G. C.

    1993-01-01

    Bicycle-related injuries have increased as cycling has become more popular. Most injuries to recreational riders are associated with overuse or improper fit of the bicycle. Injuries to racers often result from high speeds, which predispose riders to muscle strains, collisions, and falls. Cyclists contact bicycles at the pedals, seat, and handlebars. Each is associated with particular cycling injuries. Images Figure 1 Figure 3 Figure 4 Figure 5 PMID:8471908

  14. Novel driving scheme for FLCD

    NASA Astrophysics Data System (ADS)

    Wu, Jiin-chuan; Wang, Chi-Chang

    1996-03-01

    A frame change data driving scheme (FCDDS) for ferroelectric LCD(FLCD) of matrix- addressing is developed which uses only positive voltages for the row and column waveforms to achieve bipolar driving waveforms on the FLCD pixels. Thus the required supply voltage for the driver chips is half that of the conventional driving scheme. Each scan line is addressed in only twice the switching time (tau) (minimum response time of FLC) so that this scheme is suitable for high duty ratio panels. In order to meet this bistable electro-optic effect of FLCD and zero net dc voltage across each pixel of the liquid crystal, turning on and turning off pixels are done at different time slots and frame slots. This driving scheme can be easily implemented using commercially available STN LCD drivers plus a small external circuit or by making an ASIC which is a slight modification of the STN driver. Both methods are discussed.

  15. Nutrient cycling.

    PubMed

    Bormann, F H; Likens, G E

    1967-01-27

    The small-watershed approach to problems of nutrient cycling has these advantages. (i) The small watershed is a natural unit of suitable size for intensive study of nutrient cycling at the ecosystem level. (ii) It provides a means of reducing to a minimum, or virtually eliminating, the effect of the difficult-to-measure variables of geologic input and nutrient losses in deep seepage. Control of these variables makes possible accurate measurement of nutrient input and output (erosion) and therefore establishes the relationship of the smaller ecosystem to the larger biospheric cycles. (iii) The small-watershed approach provides a method whereby such important parameters as nutrient release from minerals (weathering) and annual nutrient budgets may be calculated. (iv) It provides a means of studying the interrelationships between the biota and the hydrologic cycle, various nutrient cycles, and energy flow in a single system. (v) Finally, with the small-watershed system we can test the effect of various land-management practices or environmental pollutants on nutrient cycling in natural systems. PMID:17737551

  16. On the marginal stability of upwind schemes

    NASA Astrophysics Data System (ADS)

    Gressier, J.; Moschetta, J.-M.

    Following Quirk's analysis of Roe's scheme, general criteria are derived to predict the odd-even decoupling. This analysis is applied to Roe's scheme, EFM Pullin's scheme, EIM Macrossan's scheme and AUSM Liou's scheme. Strict stability is shown to be desirable to avoid most of these flaws. Finally, the link between marginal stability and accuracy on shear waves is established.

  17. Relaxation schemes for Chebyshev spectral multigrid methods

    NASA Technical Reports Server (NTRS)

    Kang, Yimin; Fulton, Scott R.

    1993-01-01

    Two relaxation schemes for Chebyshev spectral multigrid methods are presented for elliptic equations with Dirichlet boundary conditions. The first scheme is a pointwise-preconditioned Richardson relaxation scheme and the second is a line relaxation scheme. The line relaxation scheme provides an efficient and relatively simple approach for solving two-dimensional spectral equations. Numerical examples and comparisons with other methods are given.

  18. Uncertainty of Microphysics Schemes in CRMs

    NASA Astrophysics Data System (ADS)

    Tao, W. K.; van den Heever, S. C.; Wu, D.; Saleeby, S. M.; Lang, S. E.

    2015-12-01

    Microphysics is the framework through which to understand the links between interactive aerosol, cloud and precipitation processes. These processes play a critical role in the water and energy cycle. CRMs with advanced microphysics schemes have been used to study the interaction between aerosol, cloud and precipitation processes at high resolution. But, there are still many uncertainties associated with these microphysics schemes. This has arisen, in part, from the fact microphysical processes cannot be measured directly; instead, cloud properties, which can be measured, are and have been used to validate model results. The utilization of current and future global high-resolution models is rapidly increasing and are at what has been traditional CRM resolutions and are using microphysics schemes that were developed in traditional CRMs. A potential NASA satellite mission called the Cloud and Precipitation Processes Mission (CaPPM) is currently being planned for submission to the NASA Earth Science Decadal Survey. This mission could provide the necessary global estimates of cloud and precipitation properties with which to evaluate and improve dynamical and microphysical parameterizations and the feedbacks. In order to facilitate the development of this mission, CRM simulations have been conducted to identify microphysical processes responsible for the greatest uncertainties in CRMs. In this talk, we will present results from numerical simulations conducted using two CRMs (NU-WRF and RAMS) with different dynamics, radiation, land surface and microphysics schemes. Specifically, we will conduct sensitivity tests to examine the uncertainty of the some of the key ice processes (i.e. riming, melting, freezing and shedding) in these two-microphysics schemes. The idea is to quantify how these two different models' respond (surface rainfall and its intensity, strength of cloud drafts, LWP/IWP, convective-stratiform-anvil area distribution) to changes of these key ice

  19. NUCLEAR REACTOR

    DOEpatents

    Christy, R.F.

    1958-07-15

    A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

  20. POWER REACTOR

    DOEpatents

    Zinn, W.H.

    1958-07-01

    A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

  1. NEUTRONIC REACTORS

    DOEpatents

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  2. Catalytic reactor

    DOEpatents

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  3. Bioconversion reactor

    SciTech Connect

    McCarty, P.L.; Bachmann, A.

    1992-02-25

    A bioconversion reactor is described for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible. 7 figs.

  4. Bioconversion reactor

    DOEpatents

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  5. REACTOR CONTROL

    DOEpatents

    Fortescue, P.; Nicoll, D.

    1962-04-24

    A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

  6. Interfacing primary heat sources and cycles for thermochemical hydrogen production

    SciTech Connect

    Bowman, M.G.

    1980-01-01

    Advantages cited for hydrogen production from water by coupling thermochemical cycles with primary heat include the possibility of high efficiencies. These can be realized only if the cycle approximates the criteria required to match the characteristics of the heat source. Different types of cycles may be necessary for fission reactors, for fusion reactors or for solar furnaces. Very high temperature processes based on decomposition of gaseous H/sub 2/O or CO/sub 2/ appear impractical even for projected solar technology. Cycles based on CdO decomposition are potentially quite efficient and require isothermal heat at temperatures that may be available from solar furnaces of fusion reactors. Sulfuric acid and solid sulfate cycles are potentially useful at temperatures available from each heat source. Solid sulfate cycles offer advantages for isothermal heat sources. All cycles under development include concentration and drying steps. Novel methods for improving such operations would be beneficial.

  7. NUCLEAR REACTOR

    DOEpatents

    Young, G.

    1963-01-01

    This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

  8. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  9. Simulating deep convection with a shallow convection scheme

    NASA Astrophysics Data System (ADS)

    Hohenegger, C.; Bretherton, C. S.

    2011-03-01

    Convective processes profoundly affect the global water and energy balance of our planet but remain a challenge for global climate modeling. Here we develop and investigate the suitability of a unified convection scheme, capable of handling both shallow and deep convection, to simulate cases of tropical oceanic convection, mid-latitude continental convection, and maritime shallow convection. To that aim, we employ large-eddy simulations (LES) as a benchmark to test and refine a unified convection scheme implemented in the Single-Column Community Atmosphere Model (SCAM). Our approach is motivated by previous cloud-resolving modeling studies, which have documented the gradual transition between shallow and deep convection and its possible importance for the simulated precipitation diurnal cycle. Analysis of the LES reveals that differences between shallow and deep convection, regarding cloud-base properties as well as entrainment/detrainment rates, can be related to the evaporation of precipitation. Parameterizing such effects and accordingly modifying the University of Washington shallow convection scheme, it is found that the new unified scheme can represent both shallow and deep convection as well as tropical and continental convection. Compared to the default SCAM version, the new scheme especially improves relative humidity, cloud cover and mass flux profiles. The new unified scheme also removes the well-known too early onset and peak of convective precipitation over mid-latitude continental areas.

  10. Simulating deep convection with a shallow convection scheme

    NASA Astrophysics Data System (ADS)

    Hohenegger, C.; Bretherton, C. S.

    2011-10-01

    Convective processes profoundly affect the global water and energy balance of our planet but remain a challenge for global climate modeling. Here we develop and investigate the suitability of a unified convection scheme, capable of handling both shallow and deep convection, to simulate cases of tropical oceanic convection, mid-latitude continental convection, and maritime shallow convection. To that aim, we employ large-eddy simulations (LES) as a benchmark to test and refine a unified convection scheme implemented in the Single-column Community Atmosphere Model (SCAM). Our approach is motivated by previous cloud-resolving modeling studies, which have documented the gradual transition between shallow and deep convection and its possible importance for the simulated precipitation diurnal cycle. Analysis of the LES reveals that differences between shallow and deep convection, regarding cloud-base properties as well as entrainment/detrainment rates, can be related to the evaporation of precipitation. Parameterizing such effects and accordingly modifying the University of Washington shallow convection scheme, it is found that the new unified scheme can represent both shallow and deep convection as well as tropical and mid-latitude continental convection. Compared to the default SCAM version, the new scheme especially improves relative humidity, cloud cover and mass flux profiles. The new unified scheme also removes the well-known too early onset and peak of convective precipitation over mid-latitude continental areas.

  11. The Photochemical Cycle of Halorhodopsin

    PubMed Central

    Tittor, J.; Oesterhelt, D.; Maurer, R.; Desel, H.; Uhl, R.

    1987-01-01

    Results of experiments using flash photolysis and fast difference spectroscopy suggest an extended version of the earlier published scheme of the photochemical cycle of halorhodopsin. Detailed experimental verification of the suggested photocycle is given. Due to the high resolution of the time-resolved difference spectra, absolute spectra of the intermediates in the photocycle were derived, allowing the interpretation of complex kinetic absorbance changes. PMID:19431713

  12. A Scheme for Targeting Optical SETI Observations

    NASA Astrophysics Data System (ADS)

    Shostak, Seth

    2004-06-01

    In optical SETI (OSETI) experiments, it is generally assumed that signals will be deliberate, narrowly targeted beacons sent by extraterrestrial societies to large numbers of candidate star systems. If this is so, then it may be unrealistic to expect a high duty cycle for the received signal. Ergo, an advantage accrues to any OSETI scheme that realistically suggests where and when to search. In this paper, we elaborate a proposal (Castellano, Doyle, &McIntosh 2000) for selecting regions of sky for intensive optical SETI monitoring based on characteristics of our solar system that would be visible at great distance. This can enormously lessen the amount of sky that needs to be searched. In addition, this is an attractive approach for the transmitting society because it both increases the chances of reception and provides a large reduction in energy required. With good astrometric information, the transmitter need be no more powerful than an automobile tail light.

  13. A multigrid algorithm for the cell-centered finite difference scheme

    NASA Technical Reports Server (NTRS)

    Ewing, Richard E.; Shen, Jian

    1993-01-01

    In this article, we discuss a non-variational V-cycle multigrid algorithm based on the cell-centered finite difference scheme for solving a second-order elliptic problem with discontinuous coefficients. Due to the poor approximation property of piecewise constant spaces and the non-variational nature of our scheme, one step of symmetric linear smoothing in our V-cycle multigrid scheme may fail to be a contraction. Again, because of the simple structure of the piecewise constant spaces, prolongation and restriction are trivial; we save significant computation time with very promising computational results.

  14. Vapor Compression Cycle Design Program (CYCLE_D)

    National Institute of Standards and Technology Data Gateway

    SRD 49 NIST Vapor Compression Cycle Design Program (CYCLE_D) (PC database for purchase)   The CYCLE_D database package simulates the vapor compression refrigeration cycles. It is fully compatible with REFPROP 9.0 and covers the 62 single-compound refrigerants . Fluids can be used in mixtures comprising up to five components.

  15. Neutronic reactor

    DOEpatents

    Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  16. Neutronic reactor

    DOEpatents

    Wende, Charles W. J.

    1976-08-17

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

  17. Sonochemical Reactors.

    PubMed

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation. PMID:27573503

  18. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1960-09-27

    A unit assembly is described for a neutronic reactor comprising a tube and plurality of spaced parallel sandwiches in the tube extending lengthwise thereof, each sandwich including a middle plate having a central opening for plutonium and other openings for fertile material at opposite ends of the plate.

  19. NEUTRONIC REACTOR

    DOEpatents

    Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

    1959-10-27

    BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

  20. Neutronic reactor

    DOEpatents

    Carleton, John T.

    1977-01-25

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment.

  1. High resolution schemes for hyperbolic conservation laws

    NASA Technical Reports Server (NTRS)

    Harten, A.

    1983-01-01

    A class of new explicit second order accurate finite difference schemes for the computation of weak solutions of hyperbolic conservation laws is presented. These highly nonlinear schemes are obtained by applying a nonoscillatory first order accurate scheme to an appropriately modified flux function. The so-derived second order accurate schemes achieve high resolution while preserving the robustness of the original nonoscillatory first order accurate scheme. Numerical experiments are presented to demonstrate the performance of these new schemes.

  2. Practical Combinations of Light-Water Reactors and Fast-Reactors for Future Actinide Transmutation

    SciTech Connect

    Collins, Emory D; Renier, John-Paul

    2007-01-01

    Multicycle partitioning-transmutation (P-T) studies continue to show that use of existing light-water reactors (LWRs) and new advanced light-water reactors (ALWRs) can effectively transmute transuranic (TRU) actinides, enabling initiation of full actinide recycle much earlier than waiting for the development and deployment of sufficient fast reactor (FR) capacity. The combination of initial P-T cycles using LWRs/ALWRs in parallel with economic improvements to FR usage for electricity production, and a follow-on transition period in which FRs are deployed, is a practical approach to near-term closure of the nuclear fuel cycle with full actinide recycle.

  3. Advanced nuclear fuel cycles - Main challenges and strategic choices

    SciTech Connect

    Le Biez, V.; Machiels, A.; Sowder, A.

    2013-07-01

    A graphical conceptual model of the uranium fuel cycles has been developed to capture the present, anticipated, and potential (future) nuclear fuel cycle elements. The once-through cycle and plutonium recycle in fast reactors represent two basic approaches that bound classical options for nuclear fuel cycles. Chief among these other options are mono-recycling of plutonium in thermal reactors and recycling of minor actinides in fast reactors. Mono-recycling of plutonium in thermal reactors offers modest savings in natural uranium, provides an alternative approach for present-day interim management of used fuel, and offers a potential bridging technology to development and deployment of future fuel cycles. In addition to breeder reactors' obvious fuel sustainability advantages, recycling of minor actinides in fast reactors offers an attractive concept for long-term management of the wastes, but its ultimate value is uncertain in view of the added complexity in doing so,. Ultimately, there are no simple choices for nuclear fuel cycle options, as the selection of a fuel cycle option must reflect strategic criteria and priorities that vary with national policy and market perspectives. For example, fuel cycle decision-making driven primarily by national strategic interests will likely favor energy security or proliferation resistance issues, whereas decisions driven primarily by commercial or market influences will focus on economic competitiveness.

  4. Nonlinear secret image sharing scheme.

    PubMed

    Shin, Sang-Ho; Lee, Gil-Je; Yoo, Kee-Young

    2014-01-01

    Over the past decade, most of secret image sharing schemes have been proposed by using Shamir's technique. It is based on a linear combination polynomial arithmetic. Although Shamir's technique based secret image sharing schemes are efficient and scalable for various environments, there exists a security threat such as Tompa-Woll attack. Renvall and Ding proposed a new secret sharing technique based on nonlinear combination polynomial arithmetic in order to solve this threat. It is hard to apply to the secret image sharing. In this paper, we propose a (t, n)-threshold nonlinear secret image sharing scheme with steganography concept. In order to achieve a suitable and secure secret image sharing scheme, we adapt a modified LSB embedding technique with XOR Boolean algebra operation, define a new variable m, and change a range of prime p in sharing procedure. In order to evaluate efficiency and security of proposed scheme, we use the embedding capacity and PSNR. As a result of it, average value of PSNR and embedding capacity are 44.78 (dB) and 1.74t⌈log2 m⌉ bit-per-pixel (bpp), respectively. PMID:25140334

  5. Nonlinear Secret Image Sharing Scheme

    PubMed Central

    Shin, Sang-Ho; Yoo, Kee-Young

    2014-01-01

    Over the past decade, most of secret image sharing schemes have been proposed by using Shamir's technique. It is based on a linear combination polynomial arithmetic. Although Shamir's technique based secret image sharing schemes are efficient and scalable for various environments, there exists a security threat such as Tompa-Woll attack. Renvall and Ding proposed a new secret sharing technique based on nonlinear combination polynomial arithmetic in order to solve this threat. It is hard to apply to the secret image sharing. In this paper, we propose a (t, n)-threshold nonlinear secret image sharing scheme with steganography concept. In order to achieve a suitable and secure secret image sharing scheme, we adapt a modified LSB embedding technique with XOR Boolean algebra operation, define a new variable m, and change a range of prime p in sharing procedure. In order to evaluate efficiency and security of proposed scheme, we use the embedding capacity and PSNR. As a result of it, average value of PSNR and embedding capacity are 44.78 (dB) and 1.74t⌈log2⁡m⌉ bit-per-pixel (bpp), respectively. PMID:25140334

  6. Energy partitioning schemes: a dilemma.

    PubMed

    Mayer, I

    2007-01-01

    Two closely related energy partitioning schemes, in which the total energy is presented as a sum of atomic and diatomic contributions by using the "atomic decomposition of identity", are compared on the example of N,N-dimethylformamide, a simple but chemically rich molecule. Both schemes account for different intramolecular interactions, for instance they identify the weak C-H...O intramolecular interactions, but give completely different numbers. (The energy decomposition scheme based on the virial theorem is also considered.) The comparison of the two schemes resulted in a dilemma which is especially striking when these schemes are applied for molecules distorted from their equilibrium structures: one either gets numbers which are "on the chemical scale" and have quite appealing values at the equilibrium molecular geometries, but exhibiting a counter-intuitive distance dependence (the two-center energy components increase in absolute value with the increase of the interatomic distances)--or numbers with too large absolute values but "correct" distance behaviour. The problem is connected with the quick decay of the diatomic kinetic energy components. PMID:17328441

  7. Fuel cycle cost uncertainty from nuclear fuel cycle comparison

    SciTech Connect

    Li, J.; McNelis, D.; Yim, M.S.

    2013-07-01

    This paper examined the uncertainty in fuel cycle cost (FCC) calculation by considering both model and parameter uncertainty. Four different fuel cycle options were compared in the analysis including the once-through cycle (OT), the DUPIC cycle, the MOX cycle and a closed fuel cycle with fast reactors (FR). The model uncertainty was addressed by using three different FCC modeling approaches with and without the time value of money consideration. The relative ratios of FCC in comparison to OT did not change much by using different modeling approaches. This observation was consistent with the results of the sensitivity study for the discount rate. Two different sets of data with uncertainty range of unit costs were used to address the parameter uncertainty of the FCC calculation. The sensitivity study showed that the dominating contributor to the total variance of FCC is the uranium price. In general, the FCC of OT was found to be the lowest followed by FR, MOX, and DUPIC. But depending on the uranium price, the FR cycle was found to have lower FCC over OT. The reprocessing cost was also found to have a major impact on FCC.

  8. Overview of the nuclear fuel cycle

    SciTech Connect

    Leuze, R.E.

    1981-01-01

    The use of nuclear reactors to provide electrical energy has shown considerable growth since the first nuclear plant started commercial operation in the mid 1950s. Although the main purpose of this paper is to review the fuel cycle capabilities in the United States, the introduction is a brief review of the types of nuclear reactors in use and the world-wide nuclear capacity.

  9. Overview of the nuclear fuel cycle

    SciTech Connect

    Leuze, R.E.

    1982-01-01

    The use of nuclear reactors to provide electrical energy has shown considerable growth since the first nuclear plant started commercial operation in the mid 1950s. Although the main purpose of this paper is to review the fuel cycle capabilities in the United States, the introduction is a brief review of the types of nuclear reactors in use and the world-wide nuclear capacity.

  10. Dynamic Analysis of Fuel Cycle Transitioning

    SciTech Connect

    Brent Dixon; Steve Piet; David Shropshire; Gretchen Matthern

    2009-09-01

    This paper examines the time-dependent dynamics of transitioning from a once-through fuel cycle to a closed fuel cycle. The once-through system involves only Light Water Reactors (LWRs) operating on uranium oxide fuel UOX), while the closed cycle includes both LWRs and fast spectrum reactors (FRs) in either a single-tier system or two-tier fuel system. The single-tier system includes full transuranic recycle in FRs while the two-tier system adds one pass of mixed oxide uranium-plutonium (MOX U-Pu) fuel in the LWR. While the analysis primarily focuses on burner fast reactors, transuranic conversion ratios up to 1.0 are assessed and many of the findings apply to any fuel cycle transitioning from a thermal once-through system to a synergistic thermal-fast recycle system. These findings include uranium requirements for a range of nuclear electricity growth rates, the importance of back end fuel cycle facility timing and magnitude, the impact of employing a range of fast reactor conversion ratios, system sensitivity to used fuel cooling time prior to recycle, impacts on a range of waste management indicators, and projected electricity cost ranges for once-through, single-tier and two-tier systems. The study confirmed that significant waste management benefits can be realized as soon as recycling is initiated, but natural uranium savings are minimal in this century. The use of MOX in LWRs decouples the development of recycle facilities from fast reactor fielding, but also significantly delays and limits fast reactor deployment. In all cases, fast reactor deployment was significantly below than predicted by static equilibrium analyses.

  11. An intelligent robotics control scheme

    NASA Technical Reports Server (NTRS)

    Orlando, N. E.

    1984-01-01

    The problem of robot control is viewed at the level of communicating high-level commands produced by intelligent algorithms to the actuator/sensor controllers. Four topics are considered in the design of an integrated control and communications scheme for an intelligent robotic system: the use of abstraction spaces, hierarchical versus heterarchical control, distributed processing, and the interleaving of the steps of plan creation and plan execution. A scheme is presented for an n-level distributed hierarchical/heterarchical control system that effectively interleaves intelligent planning, execution, and sensory feedback. A three-level version of this scheme has been successfully implemented in the Intelligent Systems Research Lab at NASA Langley Research Center. This implementation forms the control structure for DAISIE (Distributed Artificially Intelligent System for Interacting with the Environment), a testbed system integrating AI software with robotics hardware.

  12. Menu Cycles.

    ERIC Educational Resources Information Center

    Clayton, Alfred; Almony, John

    The curriculum guide for commercial foods instruction is designed to aid the teacher in communicating the importance of menu cycles in commercial food production. It also provides information about the necessary steps in getting food from the raw form to the finished product, and then to the consumer. In addition to providing information on how to…

  13. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    SciTech Connect

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  14. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors

    NASA Astrophysics Data System (ADS)

    Recktenwald, Geoff; Deinert, Mark

    2010-03-01

    Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks.

  15. VISION: Verifiable Fuel Cycle Simulation Model

    SciTech Connect

    Jacob J. Jacobson; Abdellatif M. Yacout; Gretchen E. Matthern; Steven J. Piet; David E. Shropshire

    2009-04-01

    The nuclear fuel cycle is a very complex system that includes considerable dynamic complexity as well as detail complexity. In the nuclear power realm, there are experts and considerable research and development in nuclear fuel development, separations technology, reactor physics and waste management. What is lacking is an overall understanding of the entire nuclear fuel cycle and how the deployment of new fuel cycle technologies affects the overall performance of the fuel cycle. The Advanced Fuel Cycle Initiative’s systems analysis group is developing a dynamic simulation model, VISION, to capture the relationships, timing and delays in and among the fuel cycle components to help develop an understanding of how the overall fuel cycle works and can transition as technologies are changed. This paper is an overview of the philosophy and development strategy behind VISION. The paper includes some descriptions of the model and some examples of how to use VISION.

  16. Regenerative Heater Optimization for Steam Turbo-Generation Cycles of Generation IV Nuclear Power Plants with a Comparison of Two Concepts for the Westinghouse International Reactor Innovative and Secure (IRIS)

    SciTech Connect

    Williams, W.C.

    2002-08-01

    The intent of this study is to discuss some of the many factors involved in the development of the design and layout of a steam turbo-generation unit as part of a modular Generation IV nuclear power plant. Of the many factors involved in the design and layout, this research will cover feed water system layout and optimization issues. The research is arranged in hopes that it can be generalized to any Generation IV system which uses a steam powered turbo-generation unit. The research is done using the ORCENT-II heat balance codes and the Salisbury methodology to be reviewed herein. The Salisbury methodology is used on an original cycle design by Famiani for the Westinghouse IRIS and the effects due to parameter variation are studied. The vital parameters of the Salisbury methodology are the incremental heater surface capital cost (S) in $/ft{sup 2}, the value of incremental power (I) in $/kW, and the overall heat transfer coefficient (U) in Btu/ft{sup 2}-degrees Fahrenheit-hr. Each is varied in order to determine the effects on the cycles overall heat rate, output, as well as, the heater surface areas. The effects of each are shown. Then the methodology is then used to compare the optimized original Famiani design consisting of seven regenerative feedwater heaters with an optimized new cycle concept, INRC8, containing four regenerative heaters. The results are shown. It can be seen that a trade between the complexity of the seven stage regenerative Famiani cycle and the simplicity of the INRC8 cycle can be made. It is desired that this methodology can be used to show the ability to evaluate modularity through the value of size a complexity of the system as well as the performance. It also shows the effectiveness of the Salisbury methodology in the optimization of regenerative cycles for such an evaluation.

  17. NEUTRONIC REACTOR

    DOEpatents

    Wade, E.J.

    1958-09-16

    This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.

  18. REACTOR MONITORING

    DOEpatents

    Bugbee, S.J.; Hanson, V.F.; Babcock, D.F.

    1959-02-01

    A neutron density inonitoring means for reactors is described. According to this invention a tunnel is provided beneath and spaced from the active portion of the reactor and extends beyond the opposite faces of the activc portion. Neutron beam holes are provided between the active portion and the tunnel and open into the tunnel near the middle thereof. A carriage operates back and forth in the tunnel and is adapted to convey a neutron detector, such as an ion chamber, and position it beneath one of the neutron beam holes. This arrangement affords convenient access of neutron density measuring instruments to a location wherein direct measurement of neutron density within the piles can be made and at the same time affords ample protection to operating personnel.

  19. REACTOR UNLOADING

    DOEpatents

    Leverett, M.C.

    1958-02-18

    This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

  20. Nuclear reactor

    DOEpatents

    Wade, Elman E.

    1979-01-01

    A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

  1. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1958-08-19

    A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

  2. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-12-15

    A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

  3. Gaseous fuel reactors for power systems

    NASA Technical Reports Server (NTRS)

    Kendall, J. S.; Rodgers, R. J.

    1977-01-01

    Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

  4. NUCLEAR REACTORS

    DOEpatents

    Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.

    1961-12-01

    An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)

  5. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashby, J.W.

    1958-09-16

    ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

  6. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1961-01-24

    A core structure for neutronic reactors adapted for the propulsion of aircraft and rockets is offered. The core is designed for cooling by gaseous media, and comprises a plurality of hollow tapered tubular segments of a porous moderating material impregniated with fissionable fuel nested about a common axis. Alternate ends of the segments are joined. In operation a coolant gas passes through the porous structure and is heated.

  7. REACTOR CONTROL

    DOEpatents

    Ruano, W.J.

    1957-12-10

    This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

  8. Fuel Cycle System Analysis Handbook

    SciTech Connect

    Steven J. Piet; Brent W. Dixon; Dirk Gombert; Edward A. Hoffman; Gretchen E. Matthern; Kent A. Williams

    2009-06-01

    This Handbook aims to improve understanding and communication regarding nuclear fuel cycle options. It is intended to assist DOE, Campaign Managers, and other presenters prepare presentations and reports. When looking for information, check here. The Handbook generally includes few details of how calculations were performed, which can be found by consulting references provided to the reader. The Handbook emphasizes results in the form of graphics and diagrams, with only enough text to explain the graphic, to ensure that the messages associated with the graphic is clear, and to explain key assumptions and methods that cause the graphed results. Some of the material is new and is not found in previous reports, for example: (1) Section 3 has system-level mass flow diagrams for 0-tier (once-through), 1-tier (UOX to CR=0.50 fast reactor), and 2-tier (UOX to MOX-Pu to CR=0.50 fast reactor) scenarios - at both static and dynamic equilibrium. (2) To help inform fast reactor transuranic (TRU) conversion ratio and uranium supply behavior, section 5 provides the sustainable fast reactor growth rate as a function of TRU conversion ratio. (3) To help clarify the difference in recycling Pu, NpPu, NpPuAm, and all-TRU, section 5 provides mass fraction, gamma, and neutron emission for those four cases for MOX, heterogeneous LWR IMF (assemblies mixing IMF and UOX pins), and a CR=0.50 fast reactor. There are data for the first 10 LWR recycle passes and equilibrium. (4) Section 6 provides information on the cycle length, planned and unplanned outages, and TRU enrichment as a function of fast reactor TRU conversion ratio, as well as the dilution of TRU feedstock by uranium in making fast reactor fuel. (The recovered uranium is considered to be more pure than recovered TRU.) The latter parameter impacts the required TRU impurity limits specified by the Fuels Campaign. (5) Section 7 provides flows for an 800-tonne UOX separation plant. (6) To complement 'tornado' economic uncertainty

  9. Performance of dimming control scheme in visible light communication system.

    PubMed

    Wang, Zixiong; Zhong, Wen-De; Yu, Changyuan; Chen, Jian; Francois, Chin Po Shin; Chen, Wei

    2012-08-13

    We investigate the performance of visible light communication (VLC) system with a pulse width modulation (PWM) dimming control scheme. Under this scheme, the communication quality in terms of number of transmitted bits and bit error rate (BER) of less than 10(-3) should be guaranteed. However, for on-off-keying (OOK) signal, the required data rate becomes 10 times as high as the original data rate when the duty cycle of dimming control signal is 0.1. To make the dimming control scheme easy to be implemented in VLC system, we propose the variable M-QAM OFDM VLC system, where M is adjusted according to the brightness of LED light in terms of duty cycle. The results show that with different duty cycles the required data rates are not higher than the original value and less LED lamp power is required to guarantee the communication quality, which makes the dimming control system that satisfies both communication and illumination requirements easy to be implemented and power-saving. PMID:23038525

  10. Fundamental Limitations in Advanced LC Schemes

    SciTech Connect

    Mikhailichenko, A. A.

    2010-11-04

    Fundamental limitations in acceleration gradient, emittance, alignment and polarization in acceleration schemes are considered in application for novel schemes of acceleration, including laser-plasma and structure-based schemes. Problems for each method are underlined whenever it is possible. Main attention is paid to the scheme with a tilted laser bunch.

  11. A scheme for symmetrization verification

    NASA Astrophysics Data System (ADS)

    Sancho, Pedro

    2011-08-01

    We propose a scheme for symmetrization verification in two-particle systems, based on one-particle detection and state determination. In contrast to previous proposals, it does not follow a Hong-Ou-Mandel-type approach. Moreover, the technique can be used to generate superposition states of single particles.

  12. Invisibly Sanitizable Digital Signature Scheme

    NASA Astrophysics Data System (ADS)

    Miyazaki, Kunihiko; Hanaoka, Goichiro; Imai, Hideki

    A digital signature does not allow any alteration of the document to which it is attached. Appropriate alteration of some signed documents, however, should be allowed because there are security requirements other than the integrity of the document. In the disclosure of official information, for example, sensitive information such as personal information or national secrets is masked when an official document is sanitized so that its nonsensitive information can be disclosed when it is requested by a citizen. If this disclosure is done digitally by using the current digital signature schemes, the citizen cannot verify the disclosed information because it has been altered to prevent the leakage of sensitive information. The confidentiality of official information is thus incompatible with the integrity of that information, and this is called the digital document sanitizing problem. Conventional solutions such as content extraction signatures and digitally signed document sanitizing schemes with disclosure condition control can either let the sanitizer assign disclosure conditions or hide the number of sanitized portions. The digitally signed document sanitizing scheme we propose here is based on the aggregate signature derived from bilinear maps and can do both. Moreover, the proposed scheme can sanitize a signed document invisibly, that is, no one can distinguish whether the signed document has been sanitized or not.

  13. Geophysical Inversion Through Hierarchical Scheme

    NASA Astrophysics Data System (ADS)

    Furman, A.; Huisman, J. A.

    2010-12-01

    Geophysical investigation is a powerful tool that allows non-invasive and non-destructive mapping of subsurface states and properties. However, non-uniqueness associated with the inversion process prevents the quantitative use of these methods. One major direction researchers are going is constraining the inverse problem by hydrological observations and models. An alternative to the commonly used direct inversion methods are global optimization schemes (such as genetic algorithms and Monte Carlo Markov Chain methods). However, the major limitation here is the desired high resolution of the tomographic image, which leads to a large number of parameters and an unreasonably high computational effort when using global optimization schemes. Two innovative schemes are presented here. First, a hierarchical approach is used to reduce the computational effort for the global optimization. Solution is achieved for coarse spatial resolution, and this solution is used as the starting point for finer scheme. We show that the computational effort is reduced in this way dramatically. Second, we use a direct ERT inversion as the starting point for global optimization. In this case preliminary results show that the outcome is not necessarily beneficial, probably because of spatial mismatch between the results of the direct inversion and the true resistivity field.

  14. Multigroup Reactor Lattice Cell Calculation

    Energy Science and Technology Software Center (ESTSC)

    1990-03-01

    The Winfrith Improved Multigroup Scheme (WIMS), is a general code for reactor lattice cell calculations on a wide range of reactor systems. In particular, the code will accept rod or plate fuel geometries in either regular arrays or in clusters, and the energy group structure has been chosen primarily for thermal calculations. The basic library has been compiled with 14 fast groups, 13 resonance groups and 42 thermal groups, but the user is offered themore » choice of accurate solutions in many groups or rapid calculations in few groups. Temperature dependent thermal scattering matrices for a variety of scattering laws are available in the library for the principal moderators which include hydrogen, deuterium, graphite, beryllium and oxygen. WIMSD5 is a succesor version of WIMS-D/4.« less

  15. On symmetric and upwind TVD schemes

    NASA Technical Reports Server (NTRS)

    Yee, H. C.

    1985-01-01

    A class of explicit and implicit total variation diminishing (TVD) schemes for the compressible Euler and Navier-Stokes equations was developed. They do not generate spurious oscillations across shocks and contact discontinuities. In general, shocks can be captured within 1 to 2 grid points. For the inviscid case, these schemes are divided into upwind TVD schemes and symmetric (nonupwind) TVD schemes. The upwind TVD scheme is based on the second-order TVD scheme. The symmetric TVD scheme is a generalization of Roe's and Davis' TVD Lax-Wendroff scheme. The performance of these schemes on some viscous and inviscid airfoil steady-state calculations is investigated. The symmetric and upwind TVD schemes are compared.

  16. FBR and RBR particle bed space reactors

    SciTech Connect

    Powell, J.R.; Botts, T.E.

    1983-01-01

    Compact, high-performance nuclear reactor designs based on High-Temperature Gas Reactors (HTGRs) particulate fuel are investigated. The large surface area available with the small-diameter (approx. 500 microns) particulate fuel allows very high power densities (MW's/liter), small temperature differences between fuel and coolant (approx. 10/sup 0/K), high coolant-outlet temperatures (1500 to 3000/sup 0/K, depending on design), and fast reactor startup (approx. 2 to 3 seconds). Two reactor concepts are developed - the Fixed Bed Reactor (FBR), where the fuel particles are packed into a thin annular bed between two porous cylindrical drums, and the Rotating Bed Reactor (RBR), where the fuel particles are held inside a cold rotating (typically approx. 500 rpm) porous cylindrical drum. The FBR can operate steady-state in the closed-cycle He-cooled mode or in the open-cycle H/sub 2/-cooled mode. The RBR will operate only in the open-cycle H/sub 2/-cooled mode.

  17. Supply of enriched uranium for research reactors

    SciTech Connect

    Mueller, H.

    1997-08-01

    Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.

  18. Core design study of a supercritical light water reactor with double row fuel rods

    SciTech Connect

    Zhao, C.; Wu, H.; Cao, L.; Zheng, Y.; Yang, J.; Zhang, Y.

    2012-07-01

    An equilibrium core for supercritical light water reactor has been designed. A novel type of fuel assembly with dual rows of fuel rods between water rods is chosen and optimized to get more uniform assembly power distributions. Stainless steel is used for fuel rod cladding and structural material. Honeycomb structure filled with thermal isolation is introduced to reduce the usage of stainless steel and to keep moderator temperature below the pseudo critical temperature. Water flow scheme with ascending coolant flow in inner regions is carried out to achieve high outlet temperature. In order to enhance coolant outlet temperature, the radial power distributions needs to be as flat as possible through operation cycle. Fuel loading pattern and control rod pattern are optimized to flatten power distribution at inner regions. Axial fuel enrichment is divided into three parts to control axial power peak, which affects maximum cladding surface temperature. (authors)

  19. Analysis of an OECD/NEA high-temperature reactor benchmark

    SciTech Connect

    Hosking, J. G.; Newton, T. D.; Koeberl, O.; Morris, P.; Goluoglu, S.; Tombakoglu, T.; Colak, U.; Sartori, E.

    2006-07-01

    This paper describes analyses of the OECD/NEA HTR benchmark organized by the 'Working Party on the Scientific Issues of Reactor Systems (WPRS)', formerly the 'Working Party on the Physics of Plutonium Fuels and Innovative Fuel Cycles'. The benchmark was specifically designed to provide inter-comparisons for plutonium and thorium fuels when used in HTR systems. Calculations considering uranium fuel have also been included in the benchmark, in order to identify any increased uncertainties when using plutonium or thorium fuels. The benchmark consists of five phases, which include cell and whole-core calculations. Analysis of the benchmark has been performed by a number of international participants, who have used a range of deterministic and Monte Carlo code schemes. For each of the benchmark phases, neutronics parameters have been evaluated. Comparisons are made between the results of the benchmark participants, as well as comparisons between the predictions of the deterministic calculations and those from detailed Monte Carlo calculations. (authors)

  20. Sandia National Laboratories Medical Isotope Reactor concept.

    SciTech Connect

    Coats, Richard Lee; Dahl, James J.; Parma, Edward J., Jr.

    2010-04-01

    This report describes the Sandia National Laboratories Medical Isotope Reactor and hot cell facility concepts. The reactor proposed is designed to be capable of producing 100% of the U.S. demand for the medical isotope {sup 99}Mo. The concept is novel in that the fuel for the reactor and the targets for the {sup 99}Mo production are the same. There is no driver core required. The fuel pins that are in the reactor core are processed on a 7 to 21 day irradiation cycle. The fuel is low enriched uranium oxide enriched to less than 20% {sup 235}U. The fuel pins are approximately 1 cm in diameter and 30 to 40 cm in height, clad with Zircaloy (zirconium alloy). Approximately 90 to 150 fuel pins are arranged in the core in a water pool {approx}30 ft deep. The reactor power level is 1 to 2 MW. The reactor concept is a simple design that is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days. The fuel fabrication, reactor design and operation, and {sup 99}Mo production processing use well-developed technologies that minimize the technological and licensing risks. There are no impediments that prevent this type of reactor, along with its collocated hot cell facility, from being designed, fabricated, and licensed today.

  1. RELAP5 two-phase fluid model and numerical scheme for economic LWR system simulation

    SciTech Connect

    Ransom, V.H.; Wagner, R.J.; Trapp, J.A.

    1981-01-01

    The RELAP5 two-phase fluid model and the associated numerical scheme are summarized. The experience accrued in development of a fast running light water reactor system transient analysis code is reviewed and example of the code application are given.

  2. Development of a system model for advanced small modular reactors.

    SciTech Connect

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandia's concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  3. Modification of cumulus convection and planetary boundary layer schemes in the GRAPES global model

    NASA Astrophysics Data System (ADS)

    Liu, Kun; Chen, Qiying; Sun, Jian

    2015-10-01

    Cumulus convection is a key linkage between hydrological cycle and large-scale atmospheric circulation. Cumulus parameterization scheme is an important component in numerical weather and climate modeling studies. In the Global/Regional Assimilation and Prediction Enhanced System (GRAPES), turbulent mixing and diffusion approach is applied in its shallow convection scheme. This method overestimates the vertical transport of heat and moisture fluxes but underestimates cloud water mixing ratio over the region of stratocumulus clouds. As a result, the simulated low stratocumulus clouds are less than observations. To overcome this problem, a mass flux method is employed in the shallow convection scheme to replace the original one. Meanwhile, the deep convection scheme is adjusted correspondingly. This modification is similar to that in the US NCEP Global Forecast System (GFS), which uses the simplified Arakawa Schubert Scheme (SAS). The planetary boundary layer scheme (PBL) is also revised by considering the coupling between the PBL and stratocumulus clouds. With the modification of both the cumulus and PBL schemes, the GRAPES simulation of shallow convective heating rate becomes more reasonable; total amounts of stratocumulus clouds simulated over the eastern Pacific and their vertical structure are more consistent with observations; the underestimation of stratocumulus clouds simulated by original schemes is less severe with the revised schemes. Precipitation distribution in the tropics becomes more reasonable and spurious precipitation is effectively suppressed. The westward extension and northward movement of the western Pacific subtropical high simulated with the revised schemes are more consistent with Final Operational Global Analysis (FNL) than that simulated with the original schemes. The statistical scores for the global GRAPES forecast are generally improved with the revised schemes, especially for the simulation of geopotential height in the Northern

  4. Short time cycles of purely quantum refrigerators.

    PubMed

    Feldmann, Tova; Kosloff, Ronnie

    2012-05-01

    Four stroke Otto refrigerator cycles with no classical analog are studied. Extremely short cycle times with respect to the internal timescale of the working medium characterize these refrigerators. Therefore, these cycles are termed sudden. The sudden cycles are characterized by the stable limit cycle, which is the invariant of the global cycle propagator. During their operation the states of the working medium possess significant coherence which is not erased in the equilibration segments due to the very short time allocated. This characteristic is reflected in a difference between the energy entropy and the Von Neumann entropy of the working medium. A classification scheme for sudden refrigerators is developed allowing simple approximations for the cooling power and coefficient of performance. PMID:23004710

  5. Potassium Rankine cycle nuclear power systems for spacecraft and lunar-mass surface power

    SciTech Connect

    Holcomb, R.S.

    1992-07-01

    The potassium Rankine cycle has high potential for application to nuclear power systems for spacecraft and surface power on the moon and Mars. A substantial effort on the development of Rankine cycle space power systems was carried out in the 1960`s. That effort is summarized and the status of the technology today is presented. Space power systems coupling Rankine cycle power conversion to both the SP-100 reactor and thermionic reactors as a combined power cycle are described in the paper.

  6. Nuclear reactor

    DOEpatents

    Pennell, William E.; Rowan, William J.

    1977-01-01

    A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

  7. Nuclear Reactors. Revised.

    ERIC Educational Resources Information Center

    Hogerton, John F.

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

  8. ELECTRONUCLEAR REACTOR

    DOEpatents

    Lawrence, E.O.; McMillan, E.M.; Alvarez, L.W.

    1960-04-19

    An electronuclear reactor is described in which a very high-energy particle accelerator is employed with appropriate target structure to produce an artificially produced material in commercial quantities by nuclear transformations. The principal novelty resides in the combination of an accelerator with a target for converting the accelerator beam to copious quantities of low-energy neutrons for absorption in a lattice of fertile material and moderator. The fertile material of the lattice is converted by neutron absorption reactions to an artificially produced material, e.g., plutonium, where depleted uranium is utilized as the fertile material.

  9. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashley, J.W.

    1958-12-16

    A graphite moderator structure is described for a gas-cooled nuclear reactor having a vertical orlentation wherein the structure is physically stable with regard to dlmensional changes due to Wigner growth properties of the graphite, and leakage of coolant gas along spaces in the structure is reduced. The structure is comprised of stacks of unlform right prismatic graphite blocks positioned in layers extending in the direction of the lengths of the blocks, the adjacent end faces of the blocks being separated by pairs of tiles. The blocks and tiles have central bores which are in alignment when assembled and are provided with cooperatlng keys and keyways for physical stability.

  10. REACTOR COMPONETN

    DOEpatents

    Creutz, E.C.

    1959-10-27

    A reactor fuel element comprised of a slug of fissionable material disposed in a sheath of corrosion resistantmaterial is described. The sheath is in the form of a tubular container closed at one end and is in tight-fitting engagement with the peripheral sunface of the slug. An inner cap is insented into the open end of the sheath against the slug, which end is then bent around the inner cap and welded thereto. An outer cap is then welded around its peripheny to the bent portion of the container.

  11. Fast Spectrum Molten Salt Reactor Options

    SciTech Connect

    Gehin, Jess C; Holcomb, David Eugene; Flanagan, George F; Patton, Bruce W; Howard, Rob L; Harrison, Thomas J

    2011-07-01

    During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.

  12. Impact of actinide recycle on nuclear fuel cycle health risks

    SciTech Connect

    Michaels, G.E.

    1992-06-01

    The purpose of this background paper is to summarize what is presently known about potential impacts on the impacts on the health risk of the nuclear fuel cycle form deployment of the Advanced Liquid Metal Reactor (ALMR){sup 1} and Integral Fast Reactor (IF){sup 2} technology as an actinide burning system. In a companion paper the impact on waste repository risk is addressed in some detail. Therefore, this paper focuses on the remainder of the fuel cycle.

  13. Fuel cycle cost study with HEU and LEU fuels

    SciTech Connect

    Matos, J.E.; Freese, K.E.

    1984-01-01

    Fuel cycle costs are compared for a range of /sup 235/U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors.

  14. Uranium to Electricity: The Chemistry of the Nuclear Fuel Cycle

    ERIC Educational Resources Information Center

    Settle, Frank A.

    2009-01-01

    The nuclear fuel cycle consists of a series of industrial processes that produce fuel for the production of electricity in nuclear reactors, use the fuel to generate electricity, and subsequently manage the spent reactor fuel. While the physics and engineering of controlled fission are central to the generation of nuclear power, chemistry…

  15. Subranging scheme for SQUID sensors

    NASA Technical Reports Server (NTRS)

    Penanen, Konstantin I. (Inventor)

    2008-01-01

    A readout scheme for measuring the output from a SQUID-based sensor-array using an improved subranging architecture that includes multiple resolution channels (such as a coarse resolution channel and a fine resolution channel). The scheme employs a flux sensing circuit with a sensing coil connected in series to multiple input coils, each input coil being coupled to a corresponding SQUID detection circuit having a high-resolution SQUID device with independent linearizing feedback. A two-resolution configuration (course and fine) is illustrated with a primary SQUID detection circuit for generating a fine readout, and a secondary SQUID detection circuit for generating a course readout, both having feedback current coupled to the respective SQUID devices via feedback/modulation coils. The primary and secondary SQUID detection circuits function and derive independent feedback. Thus, the SQUID devices may be monitored independently of each other (and read simultaneously) to dramatically increase slew rates and dynamic range.

  16. Stable maps and Quot schemes

    NASA Astrophysics Data System (ADS)

    Popa, Mihnea; Roth, Mike

    2003-06-01

    In this paper we study the relationship between two different compactifications of the space of vector bundle quotients of an arbitrary vector bundle on a curve. One is Grothendieck's Quot scheme, while the other is a moduli space of stable maps to the relative Grassmannian. We establish an essentially optimal upper bound on the dimension of the two compactifications. Based on that, we prove that for an arbitrary vector bundle, the Quot schemes of quotients of large degree are irreducible and generically smooth. We precisely describe all the vector bundles for which the same thing holds in the case of the moduli spaces of stable maps. We show that there are in general no natural morphisms between the two compactifications. Finally, as an application, we obtain new cases of a conjecture on effective base point freeness for pluritheta linear series on moduli spaces of vector bundles.

  17. Transport Reactor Facility

    SciTech Connect

    Berry, D.A.; Shoemaker, S.A.

    1996-12-31

    The Morgantown Energy Technology Center (METC) is currently evaluating hot gas desulfurization (HGD)in its on-site transport reactor facility (TRF). This facility was originally constructed in the early 1980s to explore advanced gasification processes with an entrained reactor, and has recently been modified to incorporate a transport riser reactor. The TRF supports Integrated Gasification Combined Cycle (IGCC) power systems, one of METC`s advanced power generation systems. The HGD subsystem is a key developmental item in reducing the cost and increasing the efficiency of the IGCC concept. The TRF is a unique facility with high-temperature, high-pressure, and multiple reactant gas composition capability. The TRF can be configured for reacting a single flow pass of gas and solids using a variety of gases. The gas input system allows six different gas inputs to be mixed and heated before entering the reaction zones. Current configurations allow the use of air, carbon dioxide, carbon monoxide, hydrogen, hydrogen sulfide, methane, nitrogen, oxygen, steam, or any mixture of these gases. Construction plans include the addition of a coal gas input line. This line will bring hot coal gas from the existing Fluidized-Bed Gasifier (FBG) via the Modular Gas Cleanup Rig (MGCR) after filtering out particulates with ceramic candle filters. Solids can be fed either by a rotary pocket feeder or a screw feeder. Particle sizes may range from 70 to 150 micrometers. Both feeders have a hopper that can hold enough solid for fairly lengthy tests at the higher feed rates, thus eliminating the need for lockhopper transfers during operation.

  18. VISION - Verifiable Fuel Cycle Simulation of Nuclear Fuel Cycle Dynamics

    SciTech Connect

    Steven J. Piet; A. M. Yacout; J. J. Jacobson; C. Laws; G. E. Matthern; D. E. Shropshire

    2006-02-01

    The U.S. DOE Advanced Fuel Cycle Initiative’s (AFCI) fundamental objective is to provide technology options that - if implemented - would enable long-term growth of nuclear power while improving sustainability and energy security. The AFCI organization structure consists of four areas; Systems Analysis, Fuels, Separations and Transmutations. The Systems Analysis Working Group is tasked with bridging the program technical areas and providing the models, tools, and analyses required to assess the feasibility of design and deployment options and inform key decision makers. An integral part of the Systems Analysis tool set is the development of a system level model that can be used to examine the implications of the different mixes of reactors, implications of fuel reprocessing, impact of deployment technologies, as well as potential "exit" or "off ramp" approaches to phase out technologies, waste management issues and long-term repository needs. The Verifiable Fuel Cycle Simulation Model (VISION) is a computer-based simulation model that allows performing dynamic simulations of fuel cycles to quantify infrastructure requirements and identify key trade-offs between alternatives. It is based on the current AFCI system analysis tool "DYMOND-US" functionalities in addition to economics, isotopic decay, and other new functionalities. VISION is intended to serve as a broad systems analysis and study tool applicable to work conducted as part of the AFCI and Generation IV reactor development studies.

  19. A biometric signcryption scheme without bilinear pairing

    NASA Astrophysics Data System (ADS)

    Wang, Mingwen; Ren, Zhiyuan; Cai, Jun; Zheng, Wentao

    2013-03-01

    How to apply the entropy in biometrics into the encryption and remote authentication schemes to simplify the management of keys is a hot research area. Utilizing Dodis's fuzzy extractor method and Liu's original signcryption scheme, a biometric identity based signcryption scheme is proposed in this paper. The proposed scheme is more efficient than most of the previous proposed biometric signcryption schemes for that it does not need bilinear pairing computation and modular exponentiation computation which is time consuming largely. The analysis results show that under the CDH and DL hard problem assumption, the proposed scheme has the features of confidentiality and unforgeability simultaneously.

  20. Hybrid adsorptive membrane reactor

    DOEpatents

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  1. Hybrid adsorptive membrane reactor

    NASA Technical Reports Server (NTRS)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  2. Control Means for Reactor

    DOEpatents

    Manley, J. H.

    1961-06-27

    An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

  3. Nuclear Material Attractiveness: An Assessment of Material from PHWR's in a Closed Thorium Fuel Cycle

    SciTech Connect

    Sleaford, Brad W.; Ebbinghaus, B. B.; Bradley, Keith S.; Robel, Martin; Prichard, Andrew W.; Smith, Brian W.; Collins, Brian A.; Hase, Kevin R.; Jarvinen, G. D.; Ireland, J. R.; Johnson, M. W.; Bathke, Charles G.; Wallace, R. K.

    2010-06-11

    This paper examines the attractiveness of material mixtures containing special nuclear materials (SNM) associated with reprocessing and the thorium-based LWR fuel cycle. This paper expands upon the results from earlier studies [ , ] that examined the attractiveness of SNM associated with the reprocessing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR. This study shows that 233U that is produced in thorium-based fuel cycles is very attractive for weapons use. Consistent with other studies, these results also show that all fuel cycles examined to date need to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of "attractiveness levels" that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities [ ]. The methodology and key findings will be presented.

  4. NUCLEAR MATERIAL ATTRACTIVENESS: AN ASSESSMENT OF MATERIAL FROM PHWR'S IN A CLOSED THORIUM FUEL CYCLE

    SciTech Connect

    Sleaford, B W; Collins, B A; Ebbinghaus, B B; Bathke, C G; Prichard, A W; Wallace, R K; Smith, B W; Hase, K R; Bradley, K S; Robel, M; Jarvinen, G D; Ireland, J R; Johnson, M W

    2010-04-26

    This paper examines the attractiveness of material mixtures containing special nuclear materials (SNM) associated with reprocessing and the thorium-based LWR fuel cycle. This paper expands upon the results from earlier studies that examined the attractiveness of SNM associated with the reprocessing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR. This study shows that {sup 233}U that is produced in thorium-based fuel cycles is very attractive for weapons use. Consistent with other studies, these results also show that all fuel cycles examined to date need to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of 'attractiveness levels' that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities. The methodology and key findings will be presented.

  5. FUEL CYCLE POTENTIAL WASTE FOR DISPOSITION

    SciTech Connect

    Carter, J.

    2011-01-03

    The United States (U.S.) currently utilizes a once-through fuel cycle where used nuclear fuel (UNF) is stored on-site in either wet pools or in dry storage systems with ultimate disposal in a deep mined geologic repository envisioned. Within the Department of Energy's (DOE) Office of Nuclear Energy (DOE-NE), the Fuel Cycle Research and Development Program (FCR&D) develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development of advanced fuel cycles, including modified open and closed cycles. The safe management and disposition of used nuclear fuel and/or nuclear waste is a fundamental aspect of any nuclear fuel cycle. Yet, the routine disposal of used nuclear fuel and radioactive waste remains problematic. Advanced fuel cycles will generate different quantities and forms of waste than the current LWR fleet. This study analyzes the quantities and characteristics of potential waste forms including differing waste matrices, as a function of a variety of potential fuel cycle alternatives including: (1) Commercial UNF generated by uranium fuel light water reactors (LWR). Four once through fuel cycles analyzed in this study differ by varying the assumed expansion/contraction of nuclear power in the U.S. (2) Four alternative LWR used fuel recycling processes analyzed differ in the reprocessing method (aqueous vs. electro-chemical), complexity (Pu only or full transuranic (TRU) recovery) and waste forms generated. (3) Used Mixed Oxide (MOX) fuel derived from the recovered Pu utilizing a single reactor pass. (4) Potential waste forms generated by the reprocessing of fuels derived from recovered TRU utilizing multiple reactor passes.

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

    SciTech Connect

    Not Available

    1980-01-01

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

  7. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1962-12-18

    A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

  8. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1957-09-17

    A reactor of the type having coolant liquid circulated through clad fuel elements geometrically arranged in a solid moderator, such as graphite, is described. The core is enclosed in a pressure vessel and suitable shielding, wherein means is provided for circulating vapor through the core to superheat the same. This is accomplished by drawing off the liquid which has been heated in the core due to the fission of the fuel, passing it to a nozzle within a chamber where it flashes into a vapor, and then passing the vapor through separate tubes extending through the moderator to pick up more heat developed in the core due to the fission of the fuel, thereby producing superheated vapor.

  9. NEUTRONIC REACTOR

    DOEpatents

    Stewart, H.B.

    1958-12-23

    A nuclear reactor of the type speclfically designed for the irradiation of materials is discussed. In this design a central cyllndrical core of moderating material ls surrounded by an active portlon comprlsed of an annular tank contalning fissionable material immersed ln a liquid moderator. The active portion ls ln turn surrounded by a reflector, and a well ls provided in the center of the core to accommodate the materlals to be irradiated. The over-all dimensions of the core ln at least one plane are equal to or greater than twice the effective slowing down length and equal to or less than twlce the effective diffuslon length for neutrons in the core materials.

  10. Nuclear reactor

    DOEpatents

    Yant, Howard W.; Stinebiser, Karl W.; Anzur, Gregory C.

    1977-01-01

    A nuclear reactor, particularly a liquid-metal breeder reactor, whose upper internals include outlet modules for channeling the liquid-metal coolant from selected areas of the outlet of the core vertically to the outlet plenum. The modules are composed of a highly-refractory, high corrosion-resistant alloy, for example, INCONEL-718. Each module is disposed to confine and channel generally vertically the coolant emitted from a subplurality of core-component assemblies. Each module has a grid with openings, each opening disposed to receive the coolant from an assembly of the subplurality. The grid in addition serves as a holdown for the assemblies of the corresponding subplurality preventing their excessive ejection upwardly from the core. In the region directly over the core the outlet modules are of such peripheral form that they nest forming a continuum over the core-component assemblies whose outlet coolant they confine. Each subassembly includes a chimney which confines the coolant emitted by its corresponding subassemblies to generally vertical flow between the outlet of the core and the outlet plenum. Each subplurality of assemblies whose emitted coolant is confined by an outlet module includes assemblies which emit lower-temperature coolant, for example, a control-rod assembly, or fertile assemblies, and assemblies which emit coolant of substantially higher temperature, for example, fuel-rod assemblies. The coolants of different temperatures are mixed in the chimneys reducing the effect of stripping (hot-cold temperature fluctuations) on the remainder of the upper internals which are composed typically of AISI-304 or AISI-316 stainless steel.

  11. Family Development and the Family Life Cycle: An Empirical Evaluation.

    ERIC Educational Resources Information Center

    Spanier, Graham; And Others

    The concept of family life cycle has become increasingly prominent in the study of family development--the formation, maintenance, change, and dissolution of marriage and family relations. An evaluation of this concept is accomplished by examining the relationships between three possible stratification schemes: stage of the family life cycle,…

  12. EFFICIENCY PROBLEMS RELATED TO PERMANGANATE OXIDATION SCHEMES

    EPA Science Inventory

    Oxidation schemes for the in-situ destruction of chlorinated solvents, using potassium permanganate, are receiving considerable attention. Indication from field studies and from our own work are that permanganate oxidation schemes have inherent problems that could severely limit...

  13. Pay scheme preferences and health policy objectives.

    PubMed

    Abelsen, Birgit

    2011-04-01

    This paper studies the preferences among healthcare workers towards pay schemes involving different levels of risk. It identifies which pay scheme individuals would prefer for themselves, and which they think is best in furthering health policy objectives. The paper adds, methodologically, a way of defining pay schemes that include different levels of risk. A questionnaire was mailed to a random sample of 1111 dentists. Respondents provided information about their current and preferred pay schemes, and indicated which pay scheme, in their opinion, would best further overall health policy objectives. A total of 504 dentists (45%) returned the questionnaire, and there was no indication of systematic non-response bias. All public dentists had a current pay scheme based on a fixed salary and the majority of individuals preferred a pay scheme with more income risk. Their preferred pay schemes coincided with the ones believed to further stabilise healthcare personnel. The predominant current pay scheme among private dentists was based solely on individual output, and the majority of respondents preferred this pay scheme. In addition, their preferred pay schemes coincided with the ones believed to further efficiency objectives. Both public and private dentists believed that pay schemes, furthering efficiency objectives, had to include more performance-related pay than the ones believed to further stability and quality objectives. PMID:20565995

  14. Quantum Signature Scheme with Weak Arbitrator

    NASA Astrophysics Data System (ADS)

    Luo, Ming-Xing; Chen, Xiu-Bo; Yun, Deng; Yang, Yi-Xian

    2012-07-01

    In this paper, we propose one quantum signature scheme with a weak arbitrator to sign classical messages. This scheme can preserve the merits in the original arbitrated scheme with some entanglement resources, and provide a higher efficiency in transmission and reduction the complexity of implementation. The arbitrator is costless and only involved in the disagreement case.

  15. Comparative study of numerical schemes of TVD3, UNO3-ACM and optimized compact scheme

    NASA Technical Reports Server (NTRS)

    Lee, Duck-Joo; Hwang, Chang-Jeon; Ko, Duck-Kon; Kim, Jae-Wook

    1995-01-01

    Three different schemes are employed to solve the benchmark problem. The first one is a conventional TVD-MUSCL (Monotone Upwind Schemes for Conservation Laws) scheme. The second scheme is a UNO3-ACM (Uniformly Non-Oscillatory Artificial Compression Method) scheme. The third scheme is an optimized compact finite difference scheme modified by us: the 4th order Runge Kutta time stepping, the 4th order pentadiagonal compact spatial discretization with the maximum resolution characteristics. The problems of category 1 are solved by using the second (UNO3-ACM) and third (Optimized Compact) schemes. The problems of category 2 are solved by using the first (TVD3) and second (UNO3-ACM) schemes. The problem of category 5 is solved by using the first (TVD3) scheme. It can be concluded from the present calculations that the Optimized Compact scheme and the UN03-ACM show good resolutions for category 1 and category 2 respectively.

  16. Reactor and method of operation

    DOEpatents

    Wheeler, John A.

    1976-08-10

    A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

  17. Causal relationships between solar and geomagnetic cycles

    NASA Astrophysics Data System (ADS)

    Ponyavin, D. I.

    2006-12-01

    Sunspots are sui generis "hot spots" that display the most responsive regions to solar cycle changes. Rudolf Wolf in 1848 derived a simple measure of solar cyclicity by counting a number of sunspots and sunspot groups at the solar disk. Edward Sabine in 1852 announced that geomagnetic cycle was "absolutely identical" to solar cycle. However geomagnetic and sunspot indices due to their different nature do not exhibit similar variations and often manifest out of phase behavior. Long-term sunspot and geomagnetic time-series were studied using wavelet transforms and recurrence plot techniques. We have analyzed similarities and relationships between sunspot and geomagnetic cycles in order to find recurrence, synchronization and phase differences on interannual scale. Predictive schemes of the current and future solar cycles using geomagnetic proxies were analyzed and discussed.

  18. Your Menstrual Cycle

    MedlinePlus

    ... during your menstrual cycle What happens during your menstrual cycle The menstrual cycle includes not just your period, but the rise ... tool is based on a sample 28-day menstrual cycle, but every woman is different in how long ...

  19. Transuranic Waste Burning Potential of Thorium Fuel in a Fast Reactor - 12423

    SciTech Connect

    Wenner, Michael; Franceschini, Fausto; Ferroni, Paolo; Sartori, Alberto; Ricotti, Marco

    2012-07-01

    Westinghouse Electric Company (referred to as 'Westinghouse' in the rest of this paper) is proposing a 'back-to-front' approach to overcome the stalemate on nuclear waste management in the US. In this approach, requirements to further the societal acceptance of nuclear waste are such that the ultimate health hazard resulting from the waste package is 'as low as reasonably achievable'. Societal acceptability of nuclear waste can be enhanced by reducing the long-term radiotoxicity of the waste, which is currently driven primarily by the protracted radiotoxicity of the transuranic (TRU) isotopes. Therefore, a transition to a more benign radioactive waste can be accomplished by a fuel cycle capable of consuming the stockpile of TRU 'legacy' waste contained in the LWR Used Nuclear Fuel (UNF) while generating waste which is significantly less radio-toxic than that produced by the current open U-based fuel cycle (once through and variations thereof). Investigation of a fast reactor (FR) operating on a thorium-based fuel cycle, as opposed to the traditional uranium-based is performed. Due to a combination between its neutronic properties and its low position in the actinide chain, thorium not only burns the legacy TRU waste, but it does so with a minimal production of 'new' TRUs. The effectiveness of a thorium-based fast reactor to burn legacy TRU and its flexibility to incorporate various fuels and recycle schemes according to the evolving needs of the transmutation scenario have been investigated. Specifically, the potential for a high TRU burning rate, high U-233 generation rate if so desired and low concurrent production of TRU have been used as metrics for the examined cycles. Core physics simulations of a fast reactor core running on thorium-based fuels and burning an external TRU feed supply have been carried out over multiple cycles of irradiation, separation and reprocessing. The TRU burning capability as well as the core isotopic content have been characterized

  20. Medium Power Lead Alloy Fast Reactor Balance of Plant Options

    SciTech Connect

    Vaclav Dosta; Pavel Hejzlar; Neil E. Todreas; Jacopo Buongiorno

    2004-09-01

    Proper selection of the power conversion cycle is a very important step in the design of a nuclear reactor. Due to the higher core outlet temperature (~550°C) compared to that of light water reactors (~300°C), a wide portfolio of power cycles is available for the lead alloy fast reactor (LFR). Comparison of the following cycles for the LFR was performed: superheated steam (direct and indirect), supercritical steam, helium Brayton, and supercritical CO2 (S-CO2) recompression. Heat transfer from primary to secondary coolant was first analyzed and then the steam generators or heat exchangers were designed. The direct generation of steam in the lead alloy coolant was also evaluated. The resulting temperatures of the secondary fluids are in the range of 530-545°C, dictated by the fixed space available for the heat exchangers in the reactor vessel. For the direct steam generation situation, the temperature is 312°C. Optimization of each power cycle was carried out, yielding net plant efficiency of around 40% for the superheated steam cycle while the supercritical steam and S-CO2 cycles achieved net plant efficiency of 41%. The cycles were then compared based on their net plant efficiency and potential for low capital cost. The superheated steam cycle is a very good candidate cycle given its reasonably high net plant efficiency and ease of implementation based on the extensive knowledge and operating experience with this cycle. Although the supercritical steam cycle net plant efficiency is slightly better than that of the superheated steam cycle, its high complexity and high pressure result in higher capital cost, negatively affecting plant economics. The helium Brayton cycle achieves low net plant efficiency due to the low lead alloy core outlet temperature, and therefore, even though it is a simpler cycle than the steam cycles, its performance is mediocre in this application. The prime candidate, however, appears to be the S-CO2 recompression cycle, because it

  1. Software Requirements Specification Verifiable Fuel Cycle Simulation (VISION) Model

    SciTech Connect

    D. E. Shropshire; W. H. West

    2005-11-01

    The purpose of this Software Requirements Specification (SRS) is to define the top-level requirements for a Verifiable Fuel Cycle Simulation Model (VISION) of the Advanced Fuel Cycle (AFC). This simulation model is intended to serve a broad systems analysis and study tool applicable to work conducted as part of the AFCI (including costs estimates) and Generation IV reactor development studies.

  2. On-Going Comparison of Advanced Fuel Cycle Options

    SciTech Connect

    Piet, S.J.; Bennett, R.G.; Dixon, B.W.; Herring, J.S.; Shropshire, D.E.; Roth, M.; Smith, J.D.; Finck, P.; Hill, R.; Laidler, J.; Pasamehmetoglu, K.

    2004-10-03

    This paper summarizes the current comprehensive comparison of four major fuel cycle strategies: once-through, thermal recycle, thermal+fast recycle, fast recycle. It then proceeds to summarize comparison of the major technology options for the key elements of the fuel cycle that can implement each of the four strategies - separation processing, transmutation reactors, and fuels.

  3. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    NASA Astrophysics Data System (ADS)

    Bahri, Che Nor Aniza Che Zainul; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

    2015-04-01

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  4. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    SciTech Connect

    Bahri, Che Nor Aniza Che Zainul Majid, Amran Ab.; Al-Areqi, Wadeeah M.

    2015-04-29

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  5. Once-through CANDU reactor models for the ORIGEN2 computer code

    SciTech Connect

    Croff, A.G.; Bjerke, M.A.

    1980-11-01

    Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % /sup 235/U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given.

  6. A dynamic bandwidth allocation scheme for EPON

    NASA Astrophysics Data System (ADS)

    Li, Xiuyuan; Wu, Xiaojuan; Ma, Maode; Li, Wenming; Zhang, Yuanyuan

    2008-11-01

    This paper analyses current bandwidth schemes and proposes a novel dynamic bandwidth allocation scheme for EPON. According the scheme, we define four kinds of multimedia services such as Unsolicited Request Service (URS), Realtime Service (rt-S), Non-Real-time Service (nrt-S) and Best Effort (BE). Different kinds of services have different Quality of Service (QoS) requirements. Our scheme considers the diverse QoS request, e.g., delay for rt-S, throughput for nrt-S and fairness for BE. The simulation results show this novel scheme can ensure the quality of service (QoS) and improve bandwidth utilization.

  7. Reactor safety method

    DOEpatents

    Vachon, Lawrence J.

    1980-03-11

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

  8. NEUTRONIC REACTOR MANIPULATING DEVICE

    DOEpatents

    Ohlinger, L.A.

    1962-08-01

    A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)

  9. Safeguards Considerations for Thorium Fuel Cycles

    DOE PAGESBeta

    Worrall, Louise G.; Worrall, Andrew; Flanagan, George F.; Croft, Steven

    2016-04-21

    We report that by around 2025, thorium-based fuel cycles are likely to be deployed internationally. States such as China and India are pursuing research, development, and deployment pathways toward a number of commercial-scale thorium fuel cycles, and they are already building test reactors and the associated fuel cycle infrastructure. In the future, the potential exists for these emerging programs to sell, export, and deploy thorium fuel cycle technology in other states. Without technically adequate international safeguards protocols and measures in place, any future potential clandestine misuse of these fuel cycles could go undetected, compromising the deterrent value of these protocolsmore » and measures. The development of safeguards approaches for thorium-based fuel cycles is therefore a matter of some urgency. Yet, the focus of the international safeguards community remains mainly on safeguarding conventional 235U- and 239Pu-based fuel cycles while the safeguards challenges of thorium-uranium fuel cycles remain largely uninvestigated. This raises the following question: Is the International Atomic Energy Agency and international safeguards system ready for thorium fuel cycles? Furthermore, is the safeguards technology of today sufficiently mature to meet the verification challenges posed by thorium-based fuel cycles? In defining these and other related research questions, the objectives of this paper are to identify key safeguards considerations for thorium-based fuel cycles and to call for an early dialogue between the international safeguards and the nuclear fuel cycle communities to prepare for the potential safeguards challenges associated with these fuel cycles. In this paper, it is concluded that directed research and development programs are required to meet the identified safeguards challenges and to take timely action in preparation for the international deployment of thorium fuel cycles.« less

  10. Nuclear reactor

    DOEpatents

    Thomson, Wallace B.

    2004-03-16

    A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

  11. Open-ended fusion devices and reactors

    SciTech Connect

    Kawabe, T.; Nariai, H.

    1983-12-01

    Conceptual design studies on fusion reactors based upon open-ended confinement schemes, such as the tandem mirror and rf plugged cusp, have been carried out in Japan. These studies may be classified into two categories: near-term devices (Fusion Engineering Test Facility), and long-term fusion power recators. In the first category, a two-component cusp neutron source was proposed. In the second category, the GAMMA-R, a tandem-mirror power reactor, and the RFC-R, an axisymetric mirror and cusp, reactor studies are being conducted at the University of Tsukuba and the Institute of Plasma Physics. Mirror Fusion Engineering Facility parameters and a schematic are shown. The GAMMA-R central-cell design schematic is also shown.

  12. High-Order Energy Stable WENO Schemes

    NASA Technical Reports Server (NTRS)

    Yamaleev, Nail K.; Carpenter, Mark H.

    2008-01-01

    A new third-order Energy Stable Weighted Essentially NonOscillatory (ESWENO) finite difference scheme for scalar and vector linear hyperbolic equations with piecewise continuous initial conditions is developed. The new scheme is proven to be stable in the energy norm for both continuous and discontinuous solutions. In contrast to the existing high-resolution shock-capturing schemes, no assumption that the reconstruction should be total variation bounded (TVB) is explicitly required to prove stability of the new scheme. A rigorous truncation error analysis is presented showing that the accuracy of the 3rd-order ESWENO scheme is drastically improved if the tuning parameters of the weight functions satisfy certain criteria. Numerical results show that the new ESWENO scheme is stable and significantly outperforms the conventional third-order WENO finite difference scheme of Jiang and Shu in terms of accuracy, while providing essentially nonoscillatory solutions near strong discontinuities.

  13. Catalyzed deuterium fueled tandem mirror reactor assessment

    SciTech Connect

    Dobrott, D.

    1985-01-01

    This study was part of a Department of Energy supported alternate fusion fuels program at Science Applications International Corp. The purpose of this portion of the study is to perform an assessment of a conceptual tandem mirror reactor (TMR) that is fueled by the catalyzed-deuterium (Cat-d) fuel cycle with respect to the physics, technology, safety, and cost. Achievable stable betas and magnet configurations are found to be comparable for the Cat-d and d-t fueled TMR. A comparison with respect to cost, reactor performance, and technology requirements for a Cat-d fueled reactor and a comparable d-t fueled reactor such as MARS is also made.

  14. Safe new reactor for radionuclide production

    SciTech Connect

    Gray, P.L.

    1995-02-15

    In late 1995, DOE is schedule to announce a new tritium production unit. Near the end of the last NPR (New Production Reactors) program, work was directed towards eliminating risks in current designs and reducing effects of accidents. In the Heavy Water Reactor Program at Savannah River, the coolant was changed from heavy to light water. An alternative, passively safe concept uses a heavy-water-filled, zircaloy reactor calandria near the bottom of a swimming pool; the calandria is supported on a light-water-coolant inlet plenum and has upflow through assemblies in the calandria tubes. The reactor concept eliminates or reduces significantly most design basis and severe accidents that plague other deigns. The proven, current SRS tritium cycle remains intact; production within the US of medical isotopes such as Mo-99 would also be possible.

  15. Efficient implementation of weighted ENO schemes

    NASA Technical Reports Server (NTRS)

    Jiang, Guang-Shan; Shu, Chi-Wang

    1995-01-01

    In this paper, we further analyze, test, modify and improve the high order WENO (weighted essentially non-oscillatory) finite difference schemes of Liu, Osher and Chan. It was shown by Liu et al. that WENO schemes constructed from the r-th order (in L1 norm) ENO schemes are (r+1)-th order accurate. We propose a new way of measuring the smoothness of a numerical solution, emulating the idea of minimizing the total variation of the approximation, which results in a 5-th order WENO scheme for the case r = 3, instead of the 4-th order with the original smoothness measurement by Liu et al. This 5-th order WENO scheme is as fast as the 4-th order WENO scheme of Liu et al., and both schemes are about twice as fast as the 4-th order ENO schemes on vector supercomputers and as fast on serial and parallel computers. For Euler systems of gas dynamics, we suggest computing the weights from pressure and entropy instead of the characteristic values to simplify the costly characteristic procedure. The resulting WENO schemes are about twice as fast as the WENO schemes using the characteristic decompositions to compute weights, and work well for problems which do not contain strong shocks or strong reflected waves. We also prove that, for conservation laws with smooth solutions, all WENO schemes are convergent. Many numerical tests, including the 1D steady state nozzle flow problem and 2D shock entropy wave interaction problem, are presented to demonstrate the remarkable capability of the WENO schemes, especially the WENO scheme using the new smoothness measurement, in resolving complicated shock and flow structures. We have also applied Yang's artificial compression method to the WENO schemes to sharpen contact discontinuities.

  16. Hot Gas Desulfurization Using Transport Reactors

    SciTech Connect

    Moorehead, E.L.

    1996-12-31

    Sierra Pacific Power Company is building a 100 MW, IGCC power plant based on KRW fluid bed gasifier technology that utilizes transport reactors for hot gas desulfurization and sorbent regeneration. Use of a transport absorber avoids the need for pre-filtration of dust-laden gasifier effluent, while a transport regenerator allows for the use of 100% air without the need for heat exchange equipment. Selection of transport reactors for hot gas desulfurization using a proprietary sorbent, based on testing performed in a transport reactor test unit (TRTU) at the M. W. Kellogg Technology Development Center and in a fixed bed reactor at Morgantown Energy Technology Center (METC), is outlined. The results obtained in these two test facilities and reasons for selecting transport reactors for the IGCC power plant in preference to either fixed bed or fluidized bed reactors are discussed. This paper reviews the evolution of the hot gas desulfurization system designs and includes selected results on H{sub 2}S absorption and regeneration of sulfided sorbent over several absorption/regeneration cycles conducted in the TRTU and the METC fixed bed reactor. The original design for the Sierra Pacific Project was based on fixed bed reactors with zinc ferrite as the sorbent. Owing to the high steam requirements of this sorbent, zinc titanate was selected and tested in a fixed bed reactor and was found unacceptable due to loss of strength on cyclic absorption/regeneration operation. Another sorbent evaluated was Z-Sorb{reg_sign}, a proprietary sorbent developed by Phillips Petroleum Company, was found to have excellent sulfur capacity, structural strength and regenerability. Steam was found unsuitable as fixed bed regenerator diluent, this results in a requirement for a large amount of inert gas, whereas a transport regenerator requires no diluent. The final Sierra design features transport reactors for both desulfurization and regeneration steps using neat air. 3 refs., 3 figs., 2 tabs.

  17. Scheme of thinking quantum systems

    NASA Astrophysics Data System (ADS)

    Yukalov, V. I.; Sornette, D.

    2009-11-01

    A general approach describing quantum decision procedures is developed. The approach can be applied to quantum information processing, quantum computing, creation of artificial quantum intelligence, as well as to analyzing decision processes of human decision makers. Our basic point is to consider an active quantum system possessing its own strategic state. Processing information by such a system is analogous to the cognitive processes associated to decision making by humans. The algebra of probability operators, associated with the possible options available to the decision maker, plays the role of the algebra of observables in quantum theory of measurements. A scheme is advanced for a practical realization of decision procedures by thinking quantum systems. Such thinking quantum systems can be realized by using spin lattices, systems of magnetic molecules, cold atoms trapped in optical lattices, ensembles of quantum dots, or multilevel atomic systems interacting with electromagnetic field.

  18. Hydrological cycle.

    PubMed

    Gonçalves, H C; Mercante, M A; Santos, E T

    2011-04-01

    The Pantanal hydrological cycle holds an important meaning in the Alto Paraguay Basin, comprising two areas with considerably diverse conditions regarding natural and water resources: the Plateau and the Plains. From the perspective of the ecosystem function, the hydrological flow in the relationship between plateau and plains is important for the creation of reproductive and feeding niches for the regional biodiversity. In general, river declivity in the plateau is 0.6 m/km while declivity on the plains varies from 0.1 to 0.3 m/km. The environment in the plains is characteristically seasonal and is home to an exuberant and abundant diversity of species, including some animals threatened with extinction. When the flat surface meets the plains there is a diminished water flow on the riverbeds and, during the rainy season the rivers overflow their banks, flooding the lowlands. Average annual precipitation in the Basin is 1,396 mm, ranging from 800 mm to 1,600 mm, and the heaviest rainfall occurs in the plateau region. The low drainage capacity of the rivers and lakes that shape the Pantanal, coupled with the climate in the region, produce very high evaporation: approximately 60% of all the waters coming from the plateau are lost through evaporation. The Alto Paraguay Basin, including the Pantanal, while boasting an abundant availability of water resources, also has some spots with water scarcity in some sub-basins, at different times of the year. Climate conditions alone are not enough to explain the differences observed in the Paraguay River regime and some of its tributaries. The complexity of the hydrologic regime of the Paraguay River is due to the low declivity of the lands that comprise the Mato Grosso plains and plateau (50 to 30 cm/km from east to west and 3 to 1.5 cm/km from north to south) as well as the area's dimension, which remains periodically flooded with a large volume of water. PMID:21537597

  19. Uranium to Electricity: The Chemistry of the Nuclear Fuel Cycle

    NASA Astrophysics Data System (ADS)

    Settle, Frank A.

    2009-03-01

    The nuclear fuel cycle consists of a series of industrial processes that produce fuel for the production of electricity in nuclear reactors, use the fuel to generate electricity, and subsequently manage the spent reactor fuel. While the physics and engineering of controlled fission are central to the generation of nuclear power, chemistry dominates all other aspects of the nuclear fuel cycle. An understanding of this chemistry is necessary to address the economic, environmental, safety, and proliferation issues that are essential to any substantive evaluation of nuclear power's contribution to the global energy portfolio. This article describes the role of chemistry in each component of the cycle from the metallurgy of uranium to the disposition of spent reactor fuel. It also addresses the economics of the components of the cycle and the costs of nuclear power relative to other sources of energy.

  20. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    SciTech Connect

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  1. Reactor Monitoring with Antineutrinos - A Progress Report

    NASA Astrophysics Data System (ADS)

    Bernstein, Adam

    2012-08-01

    The Reactor Safeguards regime is the name given to a set of protocols and technologies used to monitor the consumption and production of fissile materials in nuclear reactors. The Safeguards regime is administered by the International Atomic Energy Agency (IAEA), and is an essential component of the global Treaty on Nuclear Nonproliferation, recently renewed by its 189 remaining signators. (The 190th, North Korea, withdrew from the Treaty in 2003). Beginning in Russia in the 1980s, a number of researchers worldwide have experimentally demonstrated the potential of cubic meter scale antineutrino detectors for non-intrusive real-time monitoring of fissile inventories and power output of reactors. The detectors built so far have operated tens of meters from a reactor core, outside of the containment dome, largely unattended and with remote data acquisition for an entire 1.5 year reactor cycle, and have achieved levels of sensitivity to fissile content of potential interest for the IAEA safeguards regime. In this article, I will describe the unique advantages of antineutrino detectors for cooperative monitoring, consider the prospects and benefits of increasing the range of detectability for small reactors, and provide a partial survey of ongoing global research aimed at improving near-field and far field monitoring and discovery of nuclear reactors.

  2. Neutronic assessment of stringer fuel assembly design for liquid-salt-cooledvery high temperature reactor (LS-VHTR).

    SciTech Connect

    Szakaly, F. J.; Kim, T. K.; Taiwo, T. A.

    2006-09-15

    Neutronic studies of 18-pin and 36-pin stringer fuel assemblies have been performed to ascertain that core design requirements for the Liquid-Salt Cooled Very High Temperature Reactor (LS-VHTR) can be met. Parametric studies were performed to determine core characteristics required to achieve a target core cycle length of 18 months and fuel discharge burnup greater than 100 GWd/t under the constraint that the uranium enrichment be less than 20% in order to support non-proliferation goals. The studies were done using the WIMS9 lattice code and the linear reactivity model to estimate the core reactivity balance, fuel composition, and discharge burnup. The results show that the design goals can be met using a 1-batch fuel management scheme, uranium enrichment of 15% and a fuel packing fraction of 30% or greater for the 36-pin stringer fuel assembly design.

  3. A modular reactor design for in situ synchrotron x-ray investigation of atomic layer deposition processes

    NASA Astrophysics Data System (ADS)

    Klug, Jeffrey A.; Weimer, Matthew S.; Emery, Jonathan D.; Yanguas-Gil, Angel; Seifert, Sönke; Schlepütz, Christian M.; Martinson, Alex B. F.; Elam, Jeffrey W.; Hock, Adam S.; Proslier, Thomas

    2015-11-01

    Synchrotron characterization techniques provide some of the most powerful tools for the study of film structure and chemistry. The brilliance and tunability of the Advanced Photon Source allow access to scattering and spectroscopic techniques unavailable with in-house laboratory setups and provide the opportunity to probe various atomic layer deposition (ALD) processes in situ starting at the very first deposition cycle. Here, we present the design and implementation of a portable ALD instrument which possesses a modular reactor scheme that enables simple experimental switchover between various beamlines and characterization techniques. As first examples, we present in situ results for (1) X-ray surface scattering and reflectivity measurements of epitaxial ZnO ALD on sapphire, (2) grazing-incidence small angle scattering of MnO nucleation on silicon, and (3) grazing-incidence X-ray absorption spectroscopy of nucleation-regime Er2O3 ALD on amorphous ALD alumina and single crystalline sapphire.

  4. A modular reactor design for in situ synchrotron x-ray investigation of atomic layer deposition processes

    SciTech Connect

    Klug, Jeffrey A. Emery, Jonathan D.; Martinson, Alex B. F.; Proslier, Thomas; Weimer, Matthew S.; Yanguas-Gil, Angel; Elam, Jeffrey W.; Seifert, Sönke; Schlepütz, Christian M.; Hock, Adam S.

    2015-11-15

    Synchrotron characterization techniques provide some of the most powerful tools for the study of film structure and chemistry. The brilliance and tunability of the Advanced Photon Source allow access to scattering and spectroscopic techniques unavailable with in-house laboratory setups and provide the opportunity to probe various atomic layer deposition (ALD) processes in situ starting at the very first deposition cycle. Here, we present the design and implementation of a portable ALD instrument which possesses a modular reactor scheme that enables simple experimental switchover between various beamlines and characterization techniques. As first examples, we present in situ results for (1) X-ray surface scattering and reflectivity measurements of epitaxial ZnO ALD on sapphire, (2) grazing-incidence small angle scattering of MnO nucleation on silicon, and (3) grazing-incidence X-ray absorption spectroscopy of nucleation-regime Er{sub 2}O{sub 3} ALD on amorphous ALD alumina and single crystalline sapphire.

  5. A modular reactor design for in situ synchrotron X-ray investigation of atomic layer deposition processes

    SciTech Connect

    Klug, Jeffrey A.; Weimer, Matthew S.; Emery, Jonathan D.; Yanguas-Gil, Angel; Seifert, Sonke; Schleputz, Christian M.; Martinson, Alex B. F.; Elam, Jeffrey W.; Hock, Adam S.; Proslier, Thomas

    2015-11-01

    Synchrotron characterization techniques provide some of the most powerful tools for the study of film structure and chemistry. The brilliance and tunability of the Advanced Photon Source allow access to scattering and spectroscopic techniques unavailable with in-house laboratory setups and provide the opportunity to probe various atomic layer deposition (ALD) processes in situ starting at the very first deposition cycle. Here, we present the design and implementation of a portable ALD instrument which possesses a modular reactor scheme that enables simple experimental switchover between various beamlines and characterization techniques. As first examples, we present \\textit{in situ} results for 1.) X-ray surface scattering and reflectivity measurements of epitaxial ZnO ALD on sapphire, 2.) grazing-incidence small angle scattering of MnO nucleation on silicon, and 3.) grazing-incidence X-ray absorption spectroscopy of nucleation-regime Er2O3 ALD on amorphous ALD alumina and single crystalline sapphire.

  6. Fuel cycle for a fusion neutron source

    NASA Astrophysics Data System (ADS)

    Ananyev, S. S.; Spitsyn, A. V.; Kuteev, B. V.

    2015-12-01

    The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion-fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium-tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m3Pa/s, and temperature of reactor elements up to 650°C). The deuterium-tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.

  7. Fuel cycle for a fusion neutron source

    SciTech Connect

    Ananyev, S. S. Spitsyn, A. V. Kuteev, B. V.

    2015-12-15

    The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion–fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium–tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m{sup 3}Pa/s, and temperature of reactor elements up to 650°C). The deuterium–tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.

  8. Optimization and Comparison of Direct and Indirect Supercritical Carbon Dioxide Power Plant Cycles for Nuclear Applications

    SciTech Connect

    Edwin A. Harvego; Michael G. McKellar

    2011-11-01

    There have been a number of studies involving the use of gases operating in the supercritical mode for power production and process heat applications. Supercritical carbon dioxide (CO2) is particularly attractive because it is capable of achieving relatively high power conversion cycle efficiencies in the temperature range between 550 C and 750 C. Therefore, it has the potential for use with any type of high-temperature nuclear reactor concept, assuming reactor core outlet temperatures of at least 550 C. The particular power cycle investigated in this paper is a supercritical CO2 Recompression Brayton Cycle. The CO2 Recompression Brayton Cycle can be used as either a direct or indirect power conversion cycle, depending on the reactor type and reactor outlet temperature. The advantage of this cycle when compared to the helium Brayton cycle is the lower required operating temperature; 550 C versus 850 C. However, the supercritical CO2 Recompression Brayton Cycle requires an operating pressure in the range of 20 MPa, which is considerably higher than the required helium Brayton cycle operating pressure of 8 MPa. This paper presents results of analyses performed using the UniSim process analyses software to evaluate the performance of both a direct and indirect supercritical CO2 Brayton Recompression cycle for different reactor outlet temperatures. The direct supercritical CO2 cycle transferred heat directly from a 600 MWt reactor to the supercritical CO2 working fluid supplied to the turbine generator at approximately 20 MPa. The indirect supercritical CO2 cycle assumed a helium-cooled Very High Temperature Reactor (VHTR), operating at a primary system pressure of approximately 7.0 MPa, delivered heat through an intermediate heat exchanger to the secondary indirect supercritical CO2 Brayton Recompression cycle, again operating at a pressure of about 20 MPa. For both the direct and indirect cycles, sensitivity calculations were performed for reactor outlet temperature

  9. Matroids and quantum-secret-sharing schemes

    SciTech Connect

    Sarvepalli, Pradeep; Raussendorf, Robert

    2010-05-15

    A secret-sharing scheme is a cryptographic protocol to distribute a secret state in an encoded form among a group of players such that only authorized subsets of the players can reconstruct the secret. Classically, efficient secret-sharing schemes have been shown to be induced by matroids. Furthermore, access structures of such schemes can be characterized by an excluded minor relation. No such relations are known for quantum secret-sharing schemes. In this paper we take the first steps toward a matroidal characterization of quantum-secret-sharing schemes. In addition to providing a new perspective on quantum-secret-sharing schemes, this characterization has important benefits. While previous work has shown how to construct quantum-secret-sharing schemes for general access structures, these schemes are not claimed to be efficient. In this context the present results prove to be useful; they enable us to construct efficient quantum-secret-sharing schemes for many general access structures. More precisely, we show that an identically self-dual matroid that is representable over a finite field induces a pure-state quantum-secret-sharing scheme with information rate 1.

  10. CROSS-DISCIPLINARY PHYSICS AND RELATED AREAS OF SCIENCE AND TECHNOLOGY: A Novel Filter Scheme of Data Processing for SQUID-Based Magnetocardiogram

    NASA Astrophysics Data System (ADS)

    Liu, Dang-Ting; Tian, Ye; Ren, Yu-Feng; Yu, Hong-Wei; Zhang, Li-Hua; Yang, Qian-Sheng; Chen, Geng-Hua

    2008-07-01

    We present a new filter scheme for magnetocardiogram (MCG) signal processing based on the quasi-periodic characteristic of the signals. The key points of this scheme are to determine the exact numbers of data points in each cardiac cycle by using electrocardiogram (ECG) data acquired simultaneously with the MCG signal and to normalize the MCG data sequence in each cycle into an identical length. Compared with conventional filters, the scheme has the advantage of more powerful noise suppression with less signal distortion. The desire for having high quality output signals from raw MCG data acquired in a simple shielded room or even in unshielded environment may be realized with the scheme.

  11. The closed fuel cycle

    SciTech Connect

    Froment, Antoine; Gillet, Philippe

    2007-07-01

    Available in abstract form only. Full text of publication follows: The fast growth of the world's economy coupled with the need for optimizing use of natural resources, for energy security and for climate change mitigation make energy supply one of the 21. century most daring challenges. The high reliability and efficiency of nuclear energy, its competitiveness in an energy market undergoing a new oil shock are as many factors in favor of the 'renaissance' of this greenhouse gas free energy. Over 160,000 tHM of LWR1 and AGR2 Used Nuclear Fuel (UNF) have already been unloaded from the reactor cores corresponding to 7,000 tons discharged per year worldwide. By 2030, this amount could exceed 400,000 tHM and annual unloading 14,000 tHM/year. AREVA believes that closing the nuclear fuel cycle through the treatment and recycling of Used Nuclear Fuel sustains the worldwide nuclear power expansion. It is an economically sound and environmentally responsible choice, based on the preservation of natural resources through the recycling of used fuel. It furthermore provides a safe and secure management of wastes while significantly minimizing the burden left to future generations. (authors)

  12. Autonomous Throughput Improvement Scheme Using Machine Learning Algorithms for Heterogeneous Wireless Networks Aggregation

    NASA Astrophysics Data System (ADS)

    Kon, Yohsuke; Hashiguchi, Kazuki; Ito, Masato; Hasegawa, Mikio; Ishizu, Kentaro; Murakami, Homare; Harada, Hiroshi

    It is important to optimize aggregation schemes for heterogeneous wireless networks for maximizing communication throughput utilizing any available radio access networks. In the heterogeneous networks, differences of the quality of service (QoS), such as throughput, delay and packet loss rate, of the networks makes difficult to maximize the aggregation throughput. In this paper, we firstly analyze influences of such differences in QoS to the aggregation throughput, and show that it is possible to improve the throughput by adjusting the parameters of an aggregation system. Since manual parameter optimization is difficult and takes much time, we propose an autonomous parameter tuning scheme using a machine learning algorithm for the heterogeneous wireless network aggregation. We implement the proposed scheme on a heterogeneous cognitive radio network system. The results on our experimental network with network emulators show that the proposed scheme can improve the aggregation throughput better than the conventional schemes. We also evaluate the performance using public wireless network services, such as HSDPA, WiMAX and W-CDMA, and verify that the proposed scheme can improve the aggregation throughput by iterating the learning cycle even for the public wireless networks. Our experimental results show that the proposed scheme achieves twice better aggregation throughput than the conventional schemes.

  13. Evaluation of Superimposed Sequence Components of Currents based Islanding Detection Scheme during DG Interconnections

    NASA Astrophysics Data System (ADS)

    Sareen, Karan; Bhalja, Bhavesh R.; Maheshwari, Rudra Prakash

    2016-02-01

    A new islanding detection scheme for distribution network containing different types of distributed generations (DGs) is presented in this paper. The proposed scheme is based on acquiring three phase current samples for full cycle duration of each simulation case of islanding/non-islanding conditions at the point of common coupling (PCC) of the targeted DG. Afterwards, superimposed positive & negative sequence components of current are calculated and continuously compared with pre-determined threshold values. Performance of the proposed scheme has been evaluated on diversified islanding and non-islanding events which were generated by modeling standard IEEE 34-bus system using PSCAD/EMTDC software package. The proposed scheme is capable to detect islanding condition rapidly even for perfect power balance situation for both synchronous and inverter based DGs. Furthermore, it remains stable during non-islanding events such as tripping of multiple DGs and different DG interconnection operating conditions. Therefore, the proposed scheme avoids nuisance tripping during diversified non-islanding events. At the end, comparison of the proposed scheme with the existing scheme clearly indicates its advantage over the existing scheme.

  14. Assessment for advanced fuel cycle options in CANDU

    SciTech Connect

    Morreale, A.C.; Luxat, J.C.; Friedlander, Y.

    2013-07-01

    The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a driver fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.

  15. How can conceptual schemes change teaching?

    NASA Astrophysics Data System (ADS)

    Wickman, Per-Olof

    2012-03-01

    Lundqvist, Almqvist and Östman describe a teacher's manner of teaching and the possible consequences it may have for students' meaning making. In doing this the article examines a teacher's classroom practice by systematizing the teacher's transactions with the students in terms of certain conceptual schemes, namely the epistemological moves, educational philosophies and the selective traditions of this practice. In connection to their study one may ask how conceptual schemes could change teaching. This article examines how the relationship of the conceptual schemes produced by educational researchers to educational praxis has developed from the middle of the last century to today. The relationship is described as having been transformed in three steps: (1) teacher deficit and social engineering, where conceptual schemes are little acknowledged, (2) reflecting practitioners, where conceptual schemes are mangled through teacher practice to aid the choices of already knowledgeable teachers, and (3) the mangling of the conceptual schemes by researchers through practice with the purpose of revising theory.

  16. Rapid Parameterization Schemes for Aircraft Shape Optimization

    NASA Technical Reports Server (NTRS)

    Li, Wu

    2012-01-01

    A rapid shape parameterization tool called PROTEUS is developed for aircraft shape optimization. This tool can be applied directly to any aircraft geometry that has been defined in PLOT3D format, with the restriction that each aircraft component must be defined by only one data block. PROTEUS has eight types of parameterization schemes: planform, wing surface, twist, body surface, body scaling, body camber line, shifting/scaling, and linear morphing. These parametric schemes can be applied to two types of components: wing-type surfaces (e.g., wing, canard, horizontal tail, vertical tail, and pylon) and body-type surfaces (e.g., fuselage, pod, and nacelle). These schemes permit the easy setup of commonly used shape modification methods, and each customized parametric scheme can be applied to the same type of component for any configuration. This paper explains the mathematics for these parametric schemes and uses two supersonic configurations to demonstrate the application of these schemes.

  17. Enhanced Singular Wave Reactor for Surface Power

    NASA Astrophysics Data System (ADS)

    Popa-Simil, L.

    The "CANDLE" (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) also known as singular wave reactor has many significant advantages related to elimination of the need for enrichment. The use of micro-hetero structured fuel, generically called "cer-liq-mesh" will further improve burnup up to 90%. In spite it has typically large dimensions, being heavy to be transported in space, in a single piece, but because it will deliver energy in hundreds MW level for about 100 years per charge using natural Uranium or Thorium as fuel available on the planet's surface, and because it can be assembled locally becomes a very attractive option for self sustainable power cycles. The "cer-liq-mesh" fuel based singular wave reactor is smaller, less than ¼ from the size of "Candle" reactor, and has a very high burnup reducing the fuel cycle drastically. It can be transported by parts, with extremely small probability of over-unity criticality accident and be assembled to run on the surface. This represents a better option for extraterrestrial applications; in spite it requires a more complicated fuel fabrication that pays back in a simplified fuel cycle and minimum waste.

  18. Sensitivity of diurnal variation in simulated precipitation during East Asian summer monsoon to cumulus parameterization schemes

    NASA Astrophysics Data System (ADS)

    Choi, In-Jin; Jin, Emilia Kyung; Han, Ji-Young; Kim, So-Young; Kwon, Young

    2015-12-01

    The capability to simulate the diurnal variation of precipitation over East Asia region during the summertime of 2011 is investigated using five different cumulus parameterization schemes with the Weather Research and Forecasting model. A semidiurnal cycle with a 12 h interval over land and a diurnal cycle with a 24 h interval over ocean are commonly found in all simulations, consistent with the observed diurnal cycle. Two observed dominant peaks in the early morning and afternoon are reproduced in all simulations. With overestimated precipitation rate, however, the simulated afternoon peaks occur earlier than the observed peaks by 2 h for the Kain-Fritsch (KF) and Simplified Arakawa-Schubert schemes, and by 3 h for the Betts-Miller-Janjić and Tiedtke schemes. The overestimation of simulated precipitation frequency leads to amplitude and phase errors in the precipitation rate, and the early peak time of simulated precipitation intensity intensifies the phase error in the simulation over land. The KF scheme with alternative trigger function (KFtr) based on moisture advection provides slightly better results in terms of alleviating the overestimated precipitation rate and frequency and delaying the afternoon peaks. Additional sensitivity simulations based on the change of temperature perturbation in the trigger function of the KF and KFtr schemes demonstrate the afternoon peak tends to be delayed as temperature perturbation decreases, implying the significant role of convective initiation frequency in determining diurnal peaks of precipitation. Modulation of temperature perturbation alleviates the precipitation frequency bias, while it could not resolve the precipitation intensity bias.

  19. REACTOR PHYSICS MODELING OF SPENT NUCLEAR RESEARCH REACTOR FUEL FOR SNM ATTRIBUTION AND NUCLEAR FORENSICS

    SciTech Connect

    Sternat, M.; Beals, D.; Webb, R.; Nichols, T.

    2010-06-09

    Nuclear research reactors are the least safeguarded type of reactor; in some cases this may be attributed to low risk and in most cases it is due to difficulty from dynamic operation. Research reactors vary greatly in size, fuel type, enrichment, power and burnup providing a significant challenge to any standardized safeguard system. If a whole fuel assembly was interdicted, based on geometry and other traditional forensics work, one could identify the material's origin fairly accurately. If the material has been dispersed or reprocessed, in-depth reactor physics models may be used to help with the identification. Should there be a need to attribute research reactor fuel material, the Savannah River National Laboratory would perform radiochemical analysis of samples of the material as well as other non-destructive measurements. In depth reactor physics modeling would then be performed to compare to these measured results in an attempt to associate the measured results with various reactor parameters. Several reactor physics codes are being used and considered for this purpose, including: MONTEBURNS/ORIGEN/MCNP5, CINDER/MCNPX and WIMS. In attempt to identify reactor characteristics, such as time since shutdown, burnup, or power, various isotopes are used. Complexities arise when the inherent assumptions embedded in different reactor physics codes handle the isotopes differently and may quantify them to different levels of accuracy. A technical approach to modeling spent research reactor fuel begins at the assembly level upon acquiring detailed information of the reactor to be modeled. A single assembly is run using periodic boundary conditions to simulate an infinite lattice which may be repeatedly burned to produce input fuel isotopic vectors of various burnups for a core level model. A core level model will then be constructed using the assembly level results as inputs for the specific fuel shuffling pattern in an attempt to establish an equilibrium cycle. The

  20. A cascaded coding scheme for error control

    NASA Technical Reports Server (NTRS)

    Kasami, T.; Lin, S.

    1985-01-01

    A cascaded coding scheme for error control was investigated. The scheme employs a combination of hard and soft decisions in decoding. Error performance is analyzed. If the inner and outer codes are chosen properly, extremely high reliability can be attained even for a high channel bit-error-rate. Some example schemes are studied which seem to be quite suitable for satellite down-link error control.

  1. Simulation of the Australian Mobilesat signalling scheme

    NASA Technical Reports Server (NTRS)

    Rahman, Mushfiqur

    1990-01-01

    The proposed Australian Mobilesat system will provide a range of circuit switched voice/data services using the B-series satellites. The reliability of the signalling scheme between the Network Management Station (NMS) and the mobile terminal (MT) is of critical importance to the performance of the overall system. Simulation results of the performance of the signalling scheme under various channel conditions and coding schemes are presented.

  2. Quantum Authentication Scheme Based on Entanglement Swapping

    NASA Astrophysics Data System (ADS)

    Penghao, Niu; Yuan, Chen; Chong, Li

    2016-01-01

    Based on the entanglement swapping, a quantum authentication scheme with a trusted- party is proposed in this paper. With this scheme, two users can perform mutual identity authentication to confirm each other's validity. In addition, the scheme is proved to be secure under circumstances where a malicious attacker is capable of monitoring the classical and quantum channels and has the power to forge all information on the public channel.

  3. Supercritical Brayton Cycle Nuclear Power System Concepts

    SciTech Connect

    Wright, Steven A.

    2007-01-30

    Both the NASA and DOE have programs that are investigating advanced power conversion cycles for planetary surface power on the moon or Mars, and for next generation nuclear power plants on earth. The gas Brayton cycle offers many practical solutions for space nuclear power systems and was selected as the nuclear power system of choice for the NASA Prometheus project. An alternative Brayton cycle that offers high efficiency at a lower reactor coolant outlet temperature is the supercritical Brayton cycle (SCBC). The supercritical cycle is a true Brayton cycle because it uses a single phase fluid with a compressor inlet temperature that is just above the critical point of the fluid. This paper describes the use of a supercritical Brayton cycle that achieves a cycle efficiency of 26.6% with a peak coolant temperature of 750 K and for a compressor inlet temperature of 390 K. The working fluid uses a clear odorless, nontoxic refrigerant C318 perflurocarbon (C4F8) that always operates in the gas phase. This coolant was selected because it has a critical temperature and pressure of 388.38 K and 2.777 MPa. The relatively high critical temperature allows for efficient thermal radiation that keeps the radiator mass small. The SCBC achieves high efficiency because the loop design takes advantage of the non-ideal nature of the coolant equation of state just above the critical point. The lower coolant temperature means that metal fuels, uranium oxide fuels, and uranium zirconium hydride fuels with stainless steel, ferretic steel, or superalloy cladding can be used with little mass penalty or reduction in cycle efficiency. The reactor can use liquid-metal coolants and no high temperature heat exchangers need to be developed. Indirect gas cooling or perhaps even direct gas cooling can be used if the C4F8 coolant is found to be sufficiently radiation tolerant. Other fluids can also be used in the supercritical Brayton cycle including Propane (C3H8, Tcritical = 369 K) and Hexane (C6

  4. Supercritical Brayton Cycle Nuclear Power System Concepts

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.

    2007-01-01

    Both the NASA and DOE have programs that are investigating advanced power conversion cycles for planetary surface power on the moon or Mars, and for next generation nuclear power plants on earth. The gas Brayton cycle offers many practical solutions for space nuclear power systems and was selected as the nuclear power system of choice for the NASA Prometheus project. An alternative Brayton cycle that offers high efficiency at a lower reactor coolant outlet temperature is the supercritical Brayton cycle (SCBC). The supercritical cycle is a true Brayton cycle because it uses a single phase fluid with a compressor inlet temperature that is just above the critical point of the fluid. This paper describes the use of a supercritical Brayton cycle that achieves a cycle efficiency of 26.6% with a peak coolant temperature of 750 K and for a compressor inlet temperature of 390 K. The working fluid uses a clear odorless, nontoxic refrigerant C318 perflurocarbon (C4F8) that always operates in the gas phase. This coolant was selected because it has a critical temperature and pressure of 388.38 K and 2.777 MPa. The relatively high critical temperature allows for efficient thermal radiation that keeps the radiator mass small. The SCBC achieves high efficiency because the loop design takes advantage of the non-ideal nature of the coolant equation of state just above the critical point. The lower coolant temperature means that metal fuels, uranium oxide fuels, and uranium zirconium hydride fuels with stainless steel, ferretic steel, or superalloy cladding can be used with little mass penalty or reduction in cycle efficiency. The reactor can use liquid-metal coolants and no high temperature heat exchangers need to be developed. Indirect gas cooling or perhaps even direct gas cooling can be used if the C4F8 coolant is found to be sufficiently radiation tolerant. Other fluids can also be used in the supercritical Brayton cycle including Propane (C3H8, Tcritical = 369 K) and Hexane (C6

  5. Advanced reactor physics methods for heterogeneous reactor cores

    NASA Astrophysics Data System (ADS)

    Thompson, Steven A.

    To maintain the economic viability of nuclear power the industry has begun to emphasize maximizing the efficiency and output of existing nuclear power plants by using longer fuel cycles, stretch power uprates, shorter outage lengths, mixed-oxide (MOX) fuel and more aggressive operating strategies. In order to accommodate these changes, while still satisfying the peaking factor and power envelope requirements necessary to maintain safe operation, more complexity in commercial core designs have been implemented, such as an increase in the number of sub-batches and an increase in the use of both discrete and integral burnable poisons. A consequence of the increased complexity of core designs, as well as the use of MOX fuel, is an increase in the neutronic heterogeneity of the core. Such heterogeneous cores introduce challenges for the current methods that are used for reactor analysis. New methods must be developed to address these deficiencies while still maintaining the computational efficiency of existing reactor analysis methods. In this thesis, advanced core design methodologies are developed to be able to adequately analyze the highly heterogeneous core designs which are currently in use in commercial power reactors. These methodological improvements are being pursued with the goal of not sacrificing the computational efficiency which core designers require. More specifically, the PSU nodal code NEM is being updated to include an SP3 solution option, an advanced transverse leakage option, and a semi-analytical NEM solution option.

  6. Verifiable Fuel Cycle Simulation Model (VISION): A Tool for Analyzing Nuclear Fuel Cycle Futures

    SciTech Connect

    Jacob J. Jacobson; Steven J. Piet; Gretchen E. Matthern; David E. Shropshire; Robert F. Jeffers; A. M. Yacout; Tyler Schweitzer

    2010-11-01

    The nuclear fuel cycle consists of a set of complex components that are intended to work together. To support the nuclear renaissance, it is necessary to understand the impacts of changes and timing of events in any part of the fuel cycle system such as how the system would respond to each technological change, a series of which moves the fuel cycle from where it is to a postulated future state. The system analysis working group of the United States research program on advanced fuel cycles (formerly called the Advanced Fuel Cycle Initiative) is developing a dynamic simulation model, VISION, to capture the relationships, timing, and changes in and among the fuel cycle components to help develop an understanding of how the overall fuel cycle works. This paper is an overview of the philosophy and development strategy behind VISION. The paper includes some descriptions of the model components and some examples of how to use VISION. For example, VISION users can now change yearly the selection of separation or reactor technologies, the performance characteristics of those technologies, and/or the routing of material among separation and reactor types - with the model still operating on a PC in <5 min.

  7. Fuel provision for nonbreeding deuterium-tritium fusion reactors

    SciTech Connect

    Jassby, D.L.; Katsurai, M.

    1980-01-01

    Nonbreeding D-T reactors have decisive advantages in minimum size, unit cost, variety of applications, and ease of heat removal over reactors using any other fusion cycle, and significant advantages in environmental and safety characteristics over breeding D-T reactors. Considerations of relative energy production demonstrate that the most favorable source of tritium for a widely deployed system of nonbreeding D-T reactors is the very large (approx. 10 GW thermal) semi-catalyzed-deuterium (SCD), or sub-SCD reactor, where none of the escaping /sup 3/He (> 95%) or tritium (< 25%) is reinjected for burn-up. Feasibility of the ignited SCD tokamak reactor requires spatially averaged betas of 15 to 20% with a magnetic field at the TF coils of 12 to 13 Tesla.

  8. Trade study for kWe class space reactors

    NASA Astrophysics Data System (ADS)

    Bost, Donald S.

    Recent interest by NASA and other government agencies in space reactor power systems with power levels in the 1 to 100 kWe range has prompted a review of earlier space reactor programs, as well as the ongoing SP-100 program, to identify a system that will best fulfill their needs. The candidate reactor types that were reviewed are listed. They are categorized according to the method of heat removal. The five types are: conduction cooled, heat pipe cooled, liquid metal cooled, in-core thermionic and gas cooled. The UZrH moderated reactor coupled with an organic Rankine cycle power conversion system provides an attractive system for multikilowatt, long lived missions. The reactor requires a minimum development because a similar reactor has already flown and the ORC is being developed for use in the Dynamic Isotope Power System (DIPS) and on the Space Station.

  9. The CANDU Reactor System: An Appropriate Technology.

    PubMed

    Robertson, J A

    1978-02-10

    CANDU power reactors are characterized by the combination of heavy water as moderator and pressure tubes to contain the fuel and coolant. Their excellent neutron economy provides the simplicity and low costs of once-through natural-uranium fueling. Future benefits include the prospect of a near-breeder thorium fuel cycle to provide security of fuel supply without the need to develop a new reactor such as the fast breeder. These and other features make the CANDU system an appropriate technology for countries, like Canada, of intermediate economic and industrial capacity. PMID:17788102

  10. World nuclear fuel cycle requirements 1991

    SciTech Connect

    Not Available

    1991-10-10

    The nuclear fuel cycle consists of mining and milling uranium ore, processing the uranium into a form suitable for generating electricity, burning'' the fuel in nuclear reactors, and managing the resulting spent nuclear fuel. This report presents projections of domestic and foreign requirements for natural uranium and enrichment services as well as projections of discharges of spent nuclear fuel. These fuel cycle requirements are based on the forecasts of future commercial nuclear power capacity and generation published in a recent Energy Information Administration (EIA) report. Also included in this report are projections of the amount of spent fuel discharged at the end of each fuel cycle for each nuclear generating unit in the United States. The International Nuclear Model is used for calculating the projected nuclear fuel cycle requirements. 14 figs., 38 tabs.

  11. Supercritical carbon dioxide cycle control analysis.

    SciTech Connect

    Moisseytsev, A.; Sienicki, J. J.

    2011-04-11

    This report documents work carried out during FY 2008 on further investigation of control strategies for supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle energy converters. The main focus of the present work has been on investigation of the S-CO{sub 2} cycle control and behavior under conditions not covered by previous work. An important scenario which has not been previously calculated involves cycle operation for a Sodium-Cooled Fast Reactor (SFR) following a reactor scram event and the transition to the primary coolant natural circulation and decay heat removal. The Argonne National Laboratory (ANL) Plant Dynamics Code has been applied to investigate the dynamic behavior of the 96 MWe (250 MWt) Advanced Burner Test Reactor (ABTR) S-CO{sub 2} Brayton cycle following scram. The timescale for the primary sodium flowrate to coast down and the transition to natural circulation to occur was calculated with the SAS4A/SASSYS-1 computer code and found to be about 400 seconds. It is assumed that after this time, decay heat is removed by the normal ABTR shutdown heat removal system incorporating a dedicated shutdown heat removal S-CO{sub 2} pump and cooler. The ANL Plant Dynamics Code configured for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) was utilized to model the S-CO{sub 2} Brayton cycle with a decaying liquid metal coolant flow to the Pb-to-CO{sub 2} heat exchangers and temperatures reflecting the decaying core power and heat removal by the cycle. The results obtained in this manner are approximate but indicative of the cycle transient performance. The ANL Plant Dynamics Code calculations show that the S-CO{sub 2} cycle can operate for about 400 seconds following the reactor scram driven by the thermal energy stored in the reactor structures and coolant such that heat removal from the reactor exceeds the decay heat generation. Based on the results, requirements for the shutdown heat removal system may be defined

  12. Study of fueling requirements for the Engineering Test Reactor

    SciTech Connect

    Ho, S.K.; Perkins, L.J.

    1987-10-16

    An assessment of the fueling requirement for the TIBER Engineering Test Reactor is studied. The neutral shielding pellet ablation model with the inclusion of the effects of the alpha particles is used for our study. The high electron temperature in a reactor-grade plasma makes pellet penetration very difficult. The launch length has to be very large (several tens of meters) in order to avoid pellet breakage due to the low inertial strength of DT ''ice.'' The minimum repetition rate corresponding to the largest allowable pellet, is found to be about 1 Hz. A brief survey is done on the various operational and conceptual pellet injection schemes for plasma fueling. The underlying conclusion is that an alternative fueling scheme of coaxial compact-toroid plasma gun is very likely needed for effective central fueling of reactor-grade plasmas. 16 refs.

  13. Hybrid plasmachemical reactor

    SciTech Connect

    Lelevkin, V. M. Smirnova, Yu. G.; Tokarev, A. V.

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  14. Attrition reactor system

    SciTech Connect

    Scott, C.D.; Davison, B.H.

    1993-09-28

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

  15. Attrition reactor system

    SciTech Connect

    Scott, Charles D.; Davison, Brian H.

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  16. Solar Cycle 23: An Anomalous Cycle?

    NASA Astrophysics Data System (ADS)

    de Toma, G.; White, O. R.; Chapman, G. A.; Walton, S. R.; Preminger, D. G.; Cookson, A. M.

    2004-05-01

    We discuss the importance of solar cycle 23 as a magnetically simpler cycle and a variant from recent cycles. We see a significant decrease in sunspot activity in cycle 23 relative to cycle 22, but the strength of the total solar irradiance (TSI) cycle did not change significantly. The latest SOHO/VIRGO TSI time series is analyzed using new solar variability measures obtained from full-disk solar images made at the San Fernando Observatory and the MgII 280nm index. The TSI record for the period 1986 to the present is reproduced within about 130ppm RMS using only two indices representing photospheric and chromospheric sources of variability due to magnetic regions. This is in spite of the difference in magnetic activity between the two cycles. Our results show the continuing improvement in TSI measurements and surrogates containing information necessary to account for irradiance variability.

  17. The ARIES tokamak reactor study

    SciTech Connect

    Not Available

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

  18. NEUTRONIC REACTOR POWER PLANT

    DOEpatents

    Metcalf, H.E.

    1962-12-25

    This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

  19. Period meter for reactors

    DOEpatents

    Rusch, Gordon K.

    1976-01-06

    An improved log N amplifier type nuclear reactor period meter with reduced probability for noise-induced scrams is provided. With the reactor at low power levels a sampling circuit is provided to determine the reactor period by measuring the finite change in the amplitude of the log N amplifier output signal for a predetermined time period, while at high power levels, differentiation of the log N amplifier output signal provides an additional measure of the reactor period.

  20. Reactor System Transient Code.

    Energy Science and Technology Software Center (ESTSC)

    1999-07-14

    RELAP3B describes the behavior of water-cooled nuclear reactors during postulated accidents or power transients, such as large reactivity excursions, coolant losses or pump failures. The program calculates flows, mass and energy inventories, pressures, temperatures, and steam qualities along with variables associated with reactor power, reactor heat transfer, or control systems. Its versatility allows one to describe simple hydraulic systems as well as complex reactor systems.