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Sample records for fuel cycle codes

  1. Survey of nuclear fuel-cycle codes

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, C.R.; de Saussure, G.; Marable, J.H.

    1981-04-01

    A two-month survey of nuclear fuel-cycle models was undertaken. This report presents the information forthcoming from the survey. Of the nearly thirty codes reviewed in the survey, fifteen of these codes have been identified as potentially useful in fulfilling the tasks of the Nuclear Energy Analysis Division (NEAD) as defined in their FY 1981-1982 Program Plan. Six of the fifteen codes are given individual reviews. The individual reviews address such items as the funding agency, the author and organization, the date of completion of the code, adequacy of documentation, computer requirements, history of use, variables that are input and forecast, type of reactors considered, part of fuel cycle modeled and scope of the code (international or domestic, long-term or short-term, regional or national). The report recommends that the Model Evaluation Team perform an evaluation of the EUREKA uranium mining and milling code.

  2. Development of dynamic simulation code for fuel cycle fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aoki, Isao; Seki, Yasushi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan); Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

    1999-02-01

    A dynamic simulation code for fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during 2 days pulse operation cycles. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the fuel burn and the function of exhaust, purification, and supply. The processing constants of subsystem for steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using this code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

  3. Evolution of the CYCLE code for the system analysis of the nuclear fuel cycle

    Directory of Open Access Journals (Sweden)

    A.G. Kalashnikov

    2016-06-01

    Full Text Available The CYCLE code is intended to simulate mathematically the operation of a nuclear power system (NPS with thermal and fast reactors in an open or closed nuclear fuel cycle, to develop scenarios of efficient nuclear power evolution in Russia and to analyze trends in global nuclear power. The code is based on a well-known software program, WIMSD-5B, broadly used for the design of thermal reactor cells, and on a 2D multi-group software system, RZA, for the fast neutron reactor simulation. The CYCLE code was developed at IPPE in Obninsk. This paper presents a brief review of the capabilities and information on the current status of the CYCLE code. The code allows simulation of key facilities of the external fuel cycle (fuel fabrication and reprocessing facilities, SNF storage, uranium, plutonium, neptunium, americium and curium stores, RW long-term storage sites, nuclear reactors, including RBMK-1000 reactors, existing and advanced VVER reactors (using different fuel types, and fast reactors (both existing and innovative. As an important feature, the CYCLE code allows the evolution of the fuel's nuclide composition both in reactors and at the external fuel cycle phase to be considered in details. Offered as an extra option is the capability to calculate a variety of the nuclear fuel cycle cost parameters for nuclear power plants with thermal and fast reactors. For years, the code has been successfully used as part of INPRO, an international innovative nuclear reactor and fuel cycle project. The results of studies into the Russian NPS evolution scenarios were presented at Global 2011. Some other of the CYCLE-based simulation results were presented at Global 2015.

  4. Identification and Analysis of Critical Gaps in Nuclear Fuel Cycle Codes Required by the SINEMA Program

    Energy Technology Data Exchange (ETDEWEB)

    Adrian Miron; Joshua Valentine; John Christenson; Majd Hawwari; Santosh Bhatt; Mary Lou Dunzik-Gougar: Michael Lineberry

    2009-10-01

    The current state of the art in nuclear fuel cycle (NFC) modeling is an eclectic mixture of codes with various levels of applicability, flexibility, and availability. In support of the advanced fuel cycle systems analyses, especially those by the Advanced Fuel Cycle Initiative (AFCI), Unviery of Cincinnati in collaboration with Idaho State University carried out a detailed review of the existing codes describing various aspects of the nuclear fuel cycle and identified the research and development needs required for a comprehensive model of the global nuclear energy infrastructure and the associated nuclear fuel cycles. Relevant information obtained on the NFC codes was compiled into a relational database that allows easy access to various codes' properties. Additionally, the research analyzed the gaps in the NFC computer codes with respect to their potential integration into programs that perform comprehensive NFC analysis.

  5. Development of dynamic simulation code for fuel cycle of fusion reactor. 1. Single pulse operation simulation

    Energy Technology Data Exchange (ETDEWEB)

    Aoki, Isao; Seki, Yasushi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

    1997-11-01

    A dynamic simulation code for the fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during a single pulse operation. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the function of fuel burn, exhaust, purification, and supply. The processing constants of subsystem for the steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using the code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

  6. V. S. O. P. - Computer Code System for Reactor Physics and Fuel Cycle Simulation

    OpenAIRE

    Teuchert, E.; Hansen, U.; Haas, K. A.

    1980-01-01

    V .S .O .P . (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprisesneutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based onneutron flux synthesis with depletion and shut-down features, incore and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employe...

  7. V.S.O.P. (99) for WINDOWS and UNIX : computer code system for reactor physics and fuel cycle simulation

    OpenAIRE

    Rütten, H. J.; Haas, K. A.; Brockmann, H.; Ohlig, U.; Scherer, W.

    2000-01-01

    V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. The...

  8. V.S.O.P.(97) Computer Code System for Reactor Physics and Fuel Cycle Simulation : Input Manual and Comments

    OpenAIRE

    Rütten, H. J.; Haas, K. A.; Brockmann, H.; Ohlig, U.; Scherer, W.

    1998-01-01

    V.S.O.P. (97) is a computer code system for the comprehensive numerical simulation ofthe physics of thermal reactors. It implies processing ofcross sections, the setup ofthe reactor and ofthe fuel element, repeated neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to EM and to two spatial dimensi...

  9. V.S.O.P. (99/05) Computer Code System : computer code system for reactor physics and fuel cycle simulation

    OpenAIRE

    Scherer, W.; Brockmann, H.; Haas, K. A.; Rütten, H. J.

    2005-01-01

    V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. The...

  10. V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

    Energy Technology Data Exchange (ETDEWEB)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

    2010-07-15

    V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

  11. V. S. O. P. ('94) Computer Code System for Reactor Physics and Fuel Cycle Simulation

    OpenAIRE

    Teuchert, E.; Haas, K. A.; Rütten, H. J.; Brockmann, Hans; Gerwin, Helmut; Ohlig, U.; Scherer, Winfried

    1994-01-01

    V. S. O. P. ('Very Superior Old Programs) is a system of codes lurked together for the simulationof reactor life histories and temporary in-depth research. In comprises neutron cross sectionlibraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculationwith depletion and shut-down features, in-core and out-of--pile fuel management, fuel cyclecost analysis, and thermal hydraulics (at present restricted to 's). Various techniques havebeen employed to accelerat...

  12. V.S.O.P. (99) for WINDOWS and UNIX. Computer code system for reactor physics and fuel cycle simulation

    Energy Technology Data Exchange (ETDEWEB)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Scherer, W.

    2000-10-01

    V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. V.S.O.P.(99) represents the further development of V.S.O.P. (97). Compared to its precursor, the code system has been improved in many details. Major extensions have been included concerning the thermal hydraulic sections. Beyond that, the many modules of the code-system have been condensed to only 2 executables in the ''99''-release of V.S.O.P., to be comfortably handled on a WINDOWS-PC or a UNIX-computer. The necessary data input as well as the handling and book-keeping of intermediate data sets has been condensed and simplified. A 64 MB memory should be available for the execution of the code. The hard disk requirement for the executables and the basic libraries associated with the code amounts to about 7 MB. (orig.)

  13. V.S.O.P.(97) computer code system for reactor physics and fuel cycle simulation. Input manual and comments

    Energy Technology Data Exchange (ETDEWEB)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Scherer, W.

    1998-04-01

    V.S.O.P. (97) is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies processing of cross sections, the setup of the reactor and of the fuel element, repeated neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. V.S.O.P. (97) can simulate the reactor operation from the initial core towards the equilibrium core. V.S.O.P. (97) - on the basis of V.S.O.P. (94) - has been improved with regard to a more detailed treatment of the build-up and the depletion of the heavy metal isotopes. Their chains now include the minor actinides. Resonance cross sections of the lumped resonance absorbers are evaluated burnup-dependent. Beyond this, the code has been reviewed in many details, aiming at an improved precision in the computer simulation of the features of the reactors and of their fuel cycle. The code consists of about 65000 FORTRAN statements. A memory of 32 MB should be available for its use. (orig.)

  14. HTGR fuel and fuel cycle technology

    Energy Technology Data Exchange (ETDEWEB)

    Lotts, A.L.; Coobs, J.H.

    1976-08-01

    The status of fuel and fuel cycle technology for high-temperature gas-cooled reactors (HTGRs) is reviewed. The all-ceramic core of the HTGRs permits high temperatures compared with other reactors. Core outlet temperatures of 740/sup 0/C are now available for the steam cycle. For advanced HTGRs such as are required for direct-cycle power generation and for high-temperature process heat, coolant temperatures as high as 1000/sup 0/C may be expected. The paper discusses the variations of HTGR fuel designs that meet the performance requirements and the requirements of the isotopes to be used in the fuel cycle. Also discussed are the fuel cycle possibilities, which include the low-enrichment cycle, the Th-/sup 233/U cycle, and plutonium utilization in either cycle. The status of fuel and fuel cycle development is summarized.

  15. Computer code system for the R and D of nuclear fuel cycle with fast reactor. 5. Development and application of reactor analysis code system

    Energy Technology Data Exchange (ETDEWEB)

    Yokoyama, Kenji; Hazama, Taira; Chiba, Go; Ohki, Shigeo; Ishikawa, Makoto [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2002-12-01

    In the core design of fast reactors (FRs), it is very important to improve the prediction accuracy of the nuclear characteristics for both reducing cost and ensuring reliability of FR plants. A nuclear reactor analysis code system for FRs has been developed by the Japan Nuclear Cycle Development Institute (JNC). This paper describes the outline of the calculation models and methods in the system consisting of several analysis codes, such as the cell calculation code CASUP, the core calculation code TRITAC and the sensitivity analysis code SAGEP. Some examples of verification results and improvement of the design accuracy are also introduced based on the measurement data from critical assemblies, e.g, the JUPITER experiment (USA/Japan), FCA (Japan), MASURCA (France), and BFS (Russia). Furthermore, application fields and future plans, such as the development of new generation nuclear constants and applications to MA{center_dot}FP transmutation, are described. (author)

  16. Sustainability Features of Nuclear Fuel Cycle Options

    Directory of Open Access Journals (Sweden)

    Stefano Passerini

    2012-09-01

    Full Text Available The nuclear fuel cycle is the series of stages that nuclear fuel materials go through in a cradle to grave framework. The Once Through Cycle (OTC is the current fuel cycle implemented in the United States; in which an appropriate form of the fuel is irradiated through a nuclear reactor only once before it is disposed of as waste. The discharged fuel contains materials that can be suitable for use as fuel. Thus, different types of fuel recycling technologies may be introduced in order to more fully utilize the energy potential of the fuel, or reduce the environmental impacts and proliferation concerns about the discarded fuel materials. Nuclear fuel cycle systems analysis is applied in this paper to attain a better understanding of the strengths and weaknesses of fuel cycle alternatives. Through the use of the nuclear fuel cycle analysis code CAFCA (Code for Advanced Fuel Cycle Analysis, the impact of a number of recycling technologies and the associated fuel cycle options is explored in the context of the U.S. energy scenario over 100 years. Particular focus is given to the quantification of Uranium utilization, the amount of Transuranic Material (TRU generated and the economics of the different options compared to the base-line case, the OTC option. It is concluded that LWRs and the OTC are likely to dominate the nuclear energy supply system for the period considered due to limitations on availability of TRU to initiate recycling technologies. While the introduction of U-235 initiated fast reactors can accelerate their penetration of the nuclear energy system, their higher capital cost may lead to continued preference for the LWR-OTC cycle.

  17. Computer code system for the R and D of nuclear fuel cycle with fast reactor. 4. Development of an object-oriented analysis code for estimation of the material balance in the pyrochemical reprocessing process

    Energy Technology Data Exchange (ETDEWEB)

    Okamura, Nobuo; Sato, Koji [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2002-03-01

    An analysis code using the object-oriented software EX{center_dot}TD Ver.4 was developed for the estimation of material balance for the system design of the pyrochemical reprocessing plants consisting of batch processes. This code can also estimate the radioactivity balance, decay heat balance and holdup, and easily cope with the improvement of the process flow, and so on. An example of the material balance estimation under the consideration of the solvent (molten salt) recycling time is presented for the oxide electrowinning reprocessing system designed in the feasibility study of the FBR fuel cycle system. The results indicate the possibility of reduction of the vitrified waste form volume due to the extension of the recycling time of the solvent. This paper describes the outline of the code and estimation of the material balance in the oxide electrowinning reprocessing system under consideration of the solvent recycling time. (author)

  18. V.S.O.P. (99/09) Computer Code System for Reactor Physics and Fuel Cycle Simulation; Version 2009

    OpenAIRE

    Rütten, H.-J.; Haas, K. A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

    2010-01-01

    V.S.O.P.(99/ 09) represents the further development of V.S.O.P.(99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of...

  19. Nuclear fuel cycle facility accident analysis handbook

    Energy Technology Data Exchange (ETDEWEB)

    Ayer, J E; Clark, A T; Loysen, P; Ballinger, M Y; Mishima, J; Owczarski, P C; Gregory, W S; Nichols, B D

    1988-05-01

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH.

  20. Quadratic reactivity fuel cycle model

    Energy Technology Data Exchange (ETDEWEB)

    Lewins, J.D.

    1985-11-01

    For educational purposes it is highly desirable to provide simple yet realistic models for fuel cycle and fuel economy. In particular, a lumped model without recourse to detailed spatial calculations would be very helpful in providing the student with a proper understanding of the purposes of fuel cycle calculations. A teaching model for fuel cycle studies based on a lumped model assuming the summability of partial reactivities with a linear dependence of reactivity usefully illustrates fuel utilization concepts. The linear burnup model does not satisfactorily represent natural enrichment reactors. A better model, showing the trend of initial plutonium production before subsequent fuel burnup and fission product generation, is a quadratic fit. The study of M-batch cycles, reloading 1/Mth of the core at end of cycle, is now complicated by nonlinear equations. A complete account of the asymptotic cycle for any order of M-batch refueling can be given and compared with the linear model. A complete account of the transient cycle can be obtained readily in the two-batch model and this exact solution would be useful in verifying numerical marching models. It is convenient to treat the parabolic fit rho = 1 - tau/sup 2/ as a special case of the general quadratic fit rho = 1 - C/sub tau/ - (1 - C)tau/sup 2/ in suitably normalized reactivity and cycle time units. The parabolic results are given in this paper.

  1. Proliferation resistance fuel cycle technology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. S.; Ko, W. I

    1999-02-01

    The issues of dual use in nuclear technology are analysed for nuclear fuel cycle with special focus on uranium enrichment and spent fuel reprocessing which are considered as the most sensitive components in terms of vulnerability to diversion. Technical alternatives to mitigrate the vulnerability, as has been analysed in depth during the NASAP and INFCE era in the late seventies, are reviewed to characterize the DUPIC fuel cycle alternative. On the other hand, the new realities in nuclear energy including the disposition of weapon materials as a legacy of cold war are recast in an angle of nuclear proliferation resistance and safeguards with a discussion on the concept of spent fuel standard concept and its compliance with the DUPIC fuel cycle technology. (author)

  2. Cycle Average Peak Fuel Temperature Prediction Using CAPP/GAMMA+

    Energy Technology Data Exchange (ETDEWEB)

    Tak, Nam-il; Lee, Hyun Chul; Lim, Hong Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In order to obtain a cycle average maximum fuel temperature without rigorous efforts, a neutronics/thermo-fluid coupled calculation is needed with depletion capability. Recently, a CAPP/GAMMA+ coupled code system has been developed and the initial core of PMR200 was analyzed using the CAPP/GAMMA+ code system. The GAMMA+ code is a system thermo-fluid analysis code and the CAPP code is a neutronics code. The General Atomics proposed that the design limit of the fuel temperature under normal operating conditions should be a cycle-averaged maximum value. Nonetheless, the existing works of Korea Atomic Energy Research Institute (KAERI) only calculated the maximum fuel temperature at a fixed time point, e.g., the beginning of cycle (BOC) just because the calculation capability was not ready for a cycle average value. In this work, a cycle average maximum fuel temperature has been calculated using CAPP/GAMMA+ code system for the equilibrium core of PMR200. The CAPP/GAMMA+ coupled calculation was carried out for the equilibrium core of PMR 200 from BOC to EOC to obtain a cycle average peak fuel temperature. The peak fuel temperature was predicted to be 1372 .deg. C near MOC. However, the cycle average peak fuel temperature was calculated as 1181 .deg. C, which is below the design target of 1250 .deg. C.

  3. Fuel cycle for a fusion neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Ananyev, S. S., E-mail: Ananyev-SS@nrcki.ru; Spitsyn, A. V., E-mail: spitsyn-av@nrcki.ru; Kuteev, B. V., E-mail: Kuteev-BV@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

    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.

  4. Fuel cycle for a fusion neutron source

    Science.gov (United States)

    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.

  5. Modeling the Nuclear Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Jacob J. Jacobson; A. M. Yacout; G. E. Matthern; S. J. Piet; A. Moisseytsev

    2005-07-01

    The Advanced Fuel Cycle Initiative is developing a system dynamics model as part of their broad systems analysis of future nuclear energy in the United States. The model will be used to analyze and compare various proposed technology deployment scenarios. The model will also give a better understanding of the linkages between the various components of the nuclear fuel cycle that includes uranium resources, reactor number and mix, nuclear fuel type and waste management. Each of these components is tightly connected to the nuclear fuel cycle but usually analyzed in isolation of the other parts. This model will attempt to bridge these components into a single model for analysis. This work is part of a multi-national laboratory effort between Argonne National Laboratory, Idaho National Laboratory and United States Department of Energy. This paper summarizes the basics of the system dynamics model and looks at some results from the model.

  6. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2008-03-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  7. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2009-12-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  8. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert

    2007-04-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules—24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste.

  9. Hybrid reactors. [Fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Nuclear Fuel Cycle Introductory Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-02

    The nuclear fuel cycle is a complex entity, with many stages and possibilities, encompassing natural resources, energy, science, commerce, and security, involving a host of nations around the world. This overview describes the process for generating nuclear power using fissionable nuclei.

  11. Nuclear reactors and fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    fulfill its mission that is to contribute in improving the quality of life of the Brazilian people. The nuclear fuel cycle is a series of steps involved in the production and use of fuel for nuclear reactors. The Laboratories of Chemistry and Environmental Diagnosis Center, CQMA, support the demand of Nuclear Fuel Cycle Program providing chemical characterization of uranium compounds and other related materials. In this period the Research Reactor Center (CRPq) concentrated efforts on improving equipment and systems to enable the IEA-R1 research reactor to operate at higher power, increasing the capacity of radioisotopes production, samples irradiation, tests and experiments. (author)

  12. Variants of closing the nuclear fuel cycle

    Science.gov (United States)

    Andrianova, E. A.; Davidenko, V. D.; Tsibulskiy, V. F.; Tsibulskiy, S. V.

    2015-12-01

    Influence of the nuclear energy structure, the conditions of fuel burnup, and accumulation of new fissile isotopes from the raw isotopes on the main parameters of a closed fuel cycle is considered. The effects of the breeding ratio, the cooling time of the spent fuel in the external fuel cycle, and the separation of the breeding area and the fissile isotope burning area on the parameters of the fuel cycle are analyzed.

  13. Development of Vibro-packed fuel design code

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Masahiro; Sekine, Nobuyuki; Pouchon, M.A.; Miyamoto, Hiroshi [Japan Nuclear Cycle Development Inst., Tokai Works, Waste Management and Fuel Cycle Research Center, Tokai, Ibaraki (Japan); Nakajima, Yasuo [Japan Nuclear Cycle Development Inst., Heat Office, Executive Office for Policy Planning and Administration, Tokai, Ibaraki (Japan)

    2002-06-01

    In the feasibility study of a FBR cycle system, Vibro-packed fuel is considered. For the design and evaluation, three Vibro-pack-specific models are incorporated into the design code. Matthews' sintering model is used to calculate the neck growth rate between particles. The SPACON model, introduced by Botta et al. is used to calculate the effective thermal conductivity of fuel and the heat transport between fuel and cladding. A three-dimensional distinct element method is applied to analyze the mechanical properties, resulting in empirical Young's module. Experimental data will be used for the evaluation of each model, and irradiation test data will be used for evaluation of the total code. Consequently, the code reliability will be determined and potentially increased. This will constitute an important step towards fuel licensing. (author)

  14. VISION: Verifiable Fuel Cycle Simulation Model

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Transmutation Fuel Performance Code Thermal Model Verification

    Energy Technology Data Exchange (ETDEWEB)

    Gregory K. Miller; Pavel G. Medvedev

    2007-09-01

    FRAPCON fuel performance code is being modified to be able to model performance of the nuclear fuels of interest to the Global Nuclear Energy Partnership (GNEP). The present report documents the effort for verification of the FRAPCON thermal model. It was found that, with minor modifications, FRAPCON thermal model temperature calculation agrees with that of the commercial software ABAQUS (Version 6.4-4). This report outlines the methodology of the verification, code input, and calculation results.

  16. Qualification of the Darwin code for the studies of the fuel cycle relative to the boiling water reactors; Qualification du formulaire Darwin pour les etudes du cycle du combustible pour les reacteurs a eau bouillante

    Energy Technology Data Exchange (ETDEWEB)

    Allais, V

    1998-03-01

    This thesis was carried out in the framework of fuel cycles studies in partnership with COGEMA; the aim is to determine physics parameters characterising Boiling Reactor Assemblies. Those reactors Firstly distinguish themselves from Pressurised Water Reactor by the boiling of the moderator in the core and secondary by the strong neutronics heterogeneity due to complex design. The diphasic mixture formed is characterised by the void fraction parameter. The loss of information, and neutronic studies characteristics of Boiling Water Reactors led us to make preliminary studies having in view to quantify the void fraction impact on the isotopics evolution. Studies on neutronics influence of assemblies and control rods from the immediate environment allows to define the cluster size to describe. The radial description optimisation with APOLLO-2 is necessary to improve the calculation performance and to reduce the errors coming from the modelization. The following points were studied: pellet radial discretization, clustering of cells characterized by a similar behaviour, options in flux spatial calculation (interface current formalism), self-shielding optimisation (specific to each isotopes). The three dimensional modelization with CRONOS-2 and the simplified accounting of the thermohydraulics / neutronics coupling done by a procedure developed and written during this thesis, allow an evaluation of axial distribution of void fraction, power and burn-up during the irradiation. The comparison with experimental analytic results of complete assembly and pin samples dissolutions allows the qualification of this procedure and confirms the necessity to take into account the void fraction axial variation during the evolution. The application of an automatic coupling with the DARWIN cycle code will allow a precise burnup calculation to be utilized in an industrial procedure. (author)

  17. Thorium nuclear fuel cycle technology

    Energy Technology Data Exchange (ETDEWEB)

    Eom, Tae Yoon; Do, Jae Bum; Choi, Yoon Dong; Park, Kyoung Kyum; Choi, In Kyu; Lee, Jae Won; Song, Woong Sup; Kim, Heong Woo

    1998-03-01

    Since thorium produces relatively small amount of TRU elements after irradiation in the reactor, it is considered one of possible media to mix with the elements to be transmuted. Both solid and molten-salt thorium fuel cycles were investigated. Transmutation concepts being studied involved fast breeder reactor, accelerator-driven subcritical reactor, and energy amplifier with thorium. Long-lived radionuclides, especially TRU elements, could be separated from spent fuel by a pyrochemical process which is evaluated to be proliferation resistance. Pyrochemical processes of IFR, MSRE and ATW were reviewed and evaluated in detail, regarding technological feasibility, compatibility of thorium with TRU, proliferation resistance, their economy and safety. (author). 26 refs., 22 figs

  18. Fuel Cycle System Analysis Handbook

    Energy Technology Data Exchange (ETDEWEB)

    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

  19. Plutonium in an enduring fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Pillay, K.K.S.

    1998-05-01

    Nuclear fuel cycles evolved over the past five decades have allowed many nations of the world to enjoy the benefits of nuclear energy, while contributing to the sustainable consumption of the world`s energy resources. The nuclear fuel cycle for energy production suffered many traumas since the 1970s because of perceived risks of proliferation of nuclear weapons. However, the experience of the past five decades has shown that the world community is committed to safeguarding all fissile materials and continuing the use of nuclear energy resources. Decisions of a few nations to discard spent nuclear fuels in geologic formations are contrary to the goals of an enduring nuclear fuel cycle and sustainable development being pursued by the world community. The maintenance of an enduring nuclear fuel cycle is dependent on sensible management of all the resources of the fuel cycle, including spent fuels.

  20. ARC System fuel cycle analysis capability, REBUS-2

    Energy Technology Data Exchange (ETDEWEB)

    Hosteny, R.P.

    1978-10-01

    A detailed description is given of the ARC System fuel cycle modules FCI001, FCC001, FCC002, and FCC003 which form the fuel cycle analysis modules of the ARC System. These modules, in conjunction with certain other modules of the ARC System previously described in documents of this series, form the fuel cycle analysis system called REBUS-2. The physical model upon which the REBUS-2 fuel cycle modules are based and the calculational approach used in solving this model are discussed in detail. The REBUS-2 system either solves for the infinite time (i.e., equilibrium) operating conditions of a fuel recycle system under fixed fuel management conditions, or solves for the operating conditions during each of a series of explicitly specified (i.e., nonequilibrium) sequence of burn cycles. The code has the capability to adjust the fuel enrichment, the burn time, and the control poison requirements in order to satisfy user specified constraints on criticality, discharge fuel burnup, or to give the desired multiplication constant at some specified time during the reactor operation.

  1. Waste Stream Analyses for Nuclear Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    N. R. Soelberg

    2010-08-01

    A high-level study was performed in Fiscal Year 2009 for the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) Advanced Fuel Cycle Initiative (AFCI) to provide information for a range of nuclear fuel cycle options (Wigeland 2009). At that time, some fuel cycle options could not be adequately evaluated since they were not well defined and lacked sufficient information. As a result, five families of these fuel cycle options are being studied during Fiscal Year 2010 by the Systems Analysis Campaign for the DOE NE Fuel Cycle Research and Development (FCRD) program. The quality and completeness of data available to date for the fuel cycle options is insufficient to perform quantitative radioactive waste analyses using recommended metrics. This study has been limited thus far to qualitative analyses of waste streams from the candidate fuel cycle options, because quantitative data for wastes from the front end, fuel fabrication, reactor core structure, and used fuel for these options is generally not yet available.

  2. Uncertainty Analyses of Advanced Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Laurence F. Miller; J. Preston; G. Sweder; T. Anderson; S. Janson; M. Humberstone; J. MConn; J. Clark

    2008-12-12

    The Department of Energy is developing technology, experimental protocols, computational methods, systems analysis software, and many other capabilities in order to advance the nuclear power infrastructure through the Advanced Fuel Cycle Initiative (AFDI). Our project, is intended to facilitate will-informed decision making for the selection of fuel cycle options and facilities for development.

  3. The IFR modern nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Hannum, W.H.

    1991-01-01

    Nuclear power is an essential component of the world's energy supply. The IFR program, by returning to fundamentals, offers a fresh approach to closing the nuclear fuel cycle. This closed fuel cycle represents the ultimate in efficient resource utilization and environmental accountability. 35 refs., 2 tabs.

  4. External cost assessment for nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Park, Byung Heung [Korea National University of Transportation, Chungju (Korea, Republic of); Ko, Won Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-12-15

    Nuclear power is currently the second largest power supply method in Korea and the number of nuclear power plants are planned to be increased as well. However, clear management policy for spent fuels generated from nuclear power plants has not yet been established. The back-end fuel cycle, associated with nuclear material flow after nuclear reactors is a collection of technologies designed for the spent fuel management and the spent fuel management policy is closely related with the selection of a nuclear fuel cycle. Cost is an important consideration in selection of a nuclear fuel cycle and should be determined by adding external cost to private cost. Unlike the private cost, which is a direct cost, studies on the external cost are focused on nuclear reactors and not at the nuclear fuel cycle. In this research, external cost indicators applicable to nuclear fuel cycle were derived and quantified. OT (once through), DUPIC (Direct Use of PWR SF in CANDU), PWR-MOX (PWR PUREX reprocessing), and Pyro-SFR (SFR recycling with pyroprocessing) were selected as nuclear fuel cycles which could be considered for estimating external cost in Korea. Energy supply security cost, accident risk cost, and acceptance cost were defined as external cost according to precedent and estimated after analyzing approaches which have been adopted for estimating external costs on nuclear power generation.

  5. Nuclear Fusion Fuel Cycle Research Perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Hongsuk; Koo, Daeseo; Park, Jongcheol; Kim, Yeanjin [KAERI, Daejeon (Korea, Republic of); Yun, Sei-Hun [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    As a part of the International Thermonuclear Experimental Reactor (ITER) Project, we at the Korea Atomic Energy Research Institute (KAERI) and our National Fusion Research Institute (NFRI) colleagues are investigating nuclear fusion fuel cycle hardware including a nuclear fusion fuel Storage and Delivery System (SDS). To have a better knowledge of the nuclear fusion fuel cycle, we present our research efforts not only on SDS but also on the Fuel Supply System (FS), Tokamak Exhaust Processing System (TEP), Isotope Separation System (ISS), and Detritiation System (DS). To have better knowledge of the nuclear fusion fuel cycle, we presented our research efforts not only on SDS but also on the Fuel Supply System (FS), Tokamak Exhaust Processing System (TEP), Isotope Separation System (ISS), and Detritiation System (DS). Our efforts to enhance the tritium confinement will be continued for the development of cleaner nuclear fusion power plants.

  6. Globalisation of the nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Rougeau, J.-P.; Durret, L.-F.

    1995-12-31

    Three main features of the globalisation of the nuclear fuel cycle are identified and discussed. The first is an increase in the scale of the nuclear fuel cycle materials and services markets in the past 20 years. This has been accompanied by a growth in the sophistication of the fuel cycle. Secondly, the nuclear industry is now more vulnerable to outside pressures; it is no longer possible to make strategic decisions on the industry within a country solely on national considerations. Thirdly, there are changes in the decision-making process at the political, regulatory, operational and industrial level which are the consequence of global factors. (UK).

  7. International Nuclear Fuel Cycle Fact Book

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.; Patridge, M.D.

    1991-05-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECN/NEA activities reports; not reflect any one single source but frequently represent a consolidation/combination of information.

  8. Nuclear Fuel Cycle Evaluation and Real Options

    Directory of Open Access Journals (Sweden)

    L. Havlíček

    2008-01-01

    Full Text Available The first part of this paper describes the nuclear fuel cycle. It is divided into three parts. The first part, called Front-End, covers all activities connected with fuel procurement and fabrication. The middle part of the cycle includes fuel reload design activities and the operation of the fuel in the reactor. Back-End comprises all activities ensuring safe separation of spent fuel and radioactive waste from the environment. The individual stages of the fuel cycle are strongly interrelated. Overall economic optimization is very difficult. Generally, NPV is used for an economic evaluation in the nuclear fuel cycle. However the high volatility of uranium prices in the Front-End, and the large uncertainty of both economic and technical parameters in the Back-End, make the use of NPV difficult. The real option method is able to evaluate the value added by flexibility of decision making by a company under conditions of uncertainty. The possibility of applying this method to the nuclear fuel cycle evaluation is studied. 

  9. An Integrated Fuel Depletion Calculator for Fuel Cycle Options Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Erich [Univ. of Texas, Austin, TX (United States); Scopatz, Anthony [Univ. of Wisconsin, Madison, WI (United States)

    2016-04-25

    Bright-lite is a reactor modeling software developed at the University of Texas Austin to expand upon the work done with the Bright [1] reactor modeling software. Originally, bright-lite was designed to function as a standalone reactor modeling software. However, this aim was refocused t couple bright-lite with the Cyclus fuel cycle simulator [2] to make it a module for the fuel cycle simulator.

  10. Fuel burnup calculation of Ghana MNSR using ORIGEN2 and REBUS3 codes.

    Science.gov (United States)

    Abrefah, R G; Nyarko, B J B; Fletcher, J J; Akaho, E H K

    2013-10-01

    Ghana Research Reactor-1 core is to be converted from HEU fuel to LEU fuel in the near future and managing the spent nuclear fuel is very important. A fuel depletion analysis of the GHARR-1 core was performed using ORIGEN2 and REBUS3 codes to estimate the isotopic inventory at end-of-cycle in order to help in the design of an appropriate spent fuel cask. The results obtained for both codes were consistent for U-235 burnup weight percent and Pu-239 build up as a result of burnup.

  11. The nuclear fuel cycle; Le cycle du combustible nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-05-01

    After a short introduction about nuclear power in the world, fission physics and the French nuclear power plants, this brochure describes in a digest way the different steps of the nuclear fuel cycle: uranium prospecting, mining activity, processing of uranium ores and production of uranium concentrates (yellow cake), uranium chemistry (conversion of the yellow cake into uranium hexafluoride), fabrication of nuclear fuels, use of fuels, reprocessing of spent fuels (uranium, plutonium and fission products), recycling of energetic materials, and storage of radioactive wastes. (J.S.)

  12. ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

    Energy Technology Data Exchange (ETDEWEB)

    Skutnik, Steven E. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering

    2017-06-19

    processing, and depletion/decay solvers) can be self-contained into a single executable sequence. Further, to embed this capability into other software environments (such as the Cyclus fuel cycle simulator) requires that Origen’s capabilities be encapsulated into a portable, self-contained library which other codes can then call directly through function calls, thereby directly accessing the solver and data processing capabilities of Origen. Additional components relevant to this work include modernization of the reactor data libraries used by Origen for conducting nuclear fuel depletion calculations. This work has included the development of new fuel assembly lattices not previously available (such as for CANDU heavy-water reactor assemblies) as well as validation of updated lattices for light-water reactors updated to employ modern nuclear data evaluations. The CyBORG reactor analysis module as-developed under this workscope is fully capable of dynamic calculation of depleted fuel compositions from all commercial U.S. reactor assembly types as well as a number of international fuel types, including MOX, VVER, MAGNOX, and PHWR CANDU fuel assemblies. In addition, the Origen-based depletion engine allows for CyBORG to evaluate novel fuel assembly and reactor design types via creation of Origen reactor data libraries via SCALE. The establishment of this new modeling capability affords fuel cycle modelers a substantially improved ability to model dynamically-changing fuel cycle and reactor conditions, including recycled fuel compositions from fuel cycle scenarios involving material recycle into thermal-spectrum systems.

  13. Pebble bed reactor fuel cycle optimization using particle swarm algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Tavron, Barak, E-mail: btavron@bgu.ac.il [Planning, Development and Technology Division, Israel Electric Corporation Ltd., P.O. Box 10, Haifa 31000 (Israel); Shwageraus, Eugene, E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, Trumpington Street, Cambridge CB2 1PZ (United Kingdom)

    2016-10-15

    Highlights: • Particle swarm method has been developed for fuel cycle optimization of PBR reactor. • Results show uranium utilization low sensitivity to fuel and core design parameters. • Multi-zone fuel loading pattern leads to a small improvement in uranium utilization. • Thorium mixes with highly enriched uranium yields the best uranium utilization. - Abstract: Pebble bed reactors (PBR) features, such as robust thermo-mechanical fuel design and on-line continuous fueling, facilitate wide range of fuel cycle alternatives. A range off fuel pebble types, containing different amounts of fertile or fissile fuel material, may be loaded into the reactor core. Several fuel loading zones may be used since radial mixing of the pebbles was shown to be limited. This radial separation suggests the possibility to implement the “seed-blanket” concept for the utilization of fertile fuels such as thorium, and for enhancing reactor fuel utilization. In this study, the particle-swarm meta-heuristic evolutionary optimization method (PSO) has been used to find optimal fuel cycle design which yields the highest natural uranium utilization. The PSO method is known for solving efficiently complex problems with non-linear objective function, continuous or discrete parameters and complex constrains. The VSOP system of codes has been used for PBR fuel utilization calculations and MATLAB script has been used to implement the PSO algorithm. Optimization of PBR natural uranium utilization (NUU) has been carried out for 3000 MWth High Temperature Reactor design (HTR) operating on the Once Trough Then Out (OTTO) fuel management scheme, and for 400 MWth Pebble Bed Modular Reactor (PBMR) operating on the multi-pass (MEDUL) fuel management scheme. Results showed only a modest improvement in the NUU (<5%) over reference designs. Investigation of thorium fuel cases showed that the use of HEU in combination with thorium results in the most favorable reactor performance in terms of

  14. 77 FR 19278 - Informational Meeting on Nuclear Fuel Cycle Options

    Science.gov (United States)

    2012-03-30

    ... criteria or the pros and cons of any particular fuel cycle option. Opportunity for providing input on the... Informational Meeting on Nuclear Fuel Cycle Options AGENCY: Office of Fuel Cycle Technologies, Office of Nuclear Energy, Department of Energy. ACTION: Notice of meeting. SUMMARY: The Office of Fuel Cycle...

  15. Fuel cell hybrid taxi life cycle analysis

    Energy Technology Data Exchange (ETDEWEB)

    Baptista, Patricia, E-mail: patricia.baptista@ist.utl.pt [IDMEC-Instituto Superior Tecnico, Universidade Tecnica de Lisboa, Av. Rovisco Pais, 1, 1049-001 Lisboa (Portugal); Ribau, Joao; Bravo, Joao; Silva, Carla [IDMEC-Instituto Superior Tecnico, Universidade Tecnica de Lisboa, Av. Rovisco Pais, 1, 1049-001 Lisboa (Portugal); Adcock, Paul; Kells, Ashley [Intelligent Energy, Charnwood Building, HolywellPark, Ashby Road, Loughborough, LE11 3GR (United Kingdom)

    2011-09-15

    A small fleet of classic London Taxis (Black cabs) equipped with hydrogen fuel cell power systems is being prepared for demonstration during the 2012 London Olympics. This paper presents a Life Cycle Analysis for these vehicles in terms of energy consumption and CO{sub 2} emissions, focusing on the impacts of alternative vehicle technologies for the Taxi, combining the fuel life cycle (Tank-to-Wheel and Well-to-Tank) and vehicle materials Cradle-to-Grave. An internal combustion engine diesel taxi was used as the reference vehicle for the currently available technology. This is compared to battery and fuel cell vehicle configurations. Accordingly, the following energy pathways are compared: diesel, electricity and hydrogen (derived from natural gas steam reforming). Full Life Cycle Analysis, using the PCO-CENEX drive cycle, (derived from actual London Taxi drive cycles) shows that the fuel cell powered vehicle configurations have lower energy consumption (4.34 MJ/km) and CO{sub 2} emissions (235 g/km) than both the ICE Diesel (9.54 MJ/km and 738 g/km) and the battery electric vehicle (5.81 MJ/km and 269 g/km). - Highlights: > A Life Cycle Analysis of alternative vehicle technologies for the London Taxi was performed. > The hydrogen powered vehicles have the lowest energy consumption and CO{sub 2} emissions results. > A hydrogen powered solution can be a sustainable alternative in a full life cycle framework.

  16. A study on the environmental friendliness of nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. J.; Lee, B. H.; Lee, S. Y.; Lim, C. Y.; Choi, Y. S.; Lee, Y. E.; Hong, D. S.; Cheong, J. H; Park, J. B.; Kim, K. K.; Cheong, H. Y; Song, M. C; Lee, H. J. [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of)

    1998-01-01

    The purpose of this study is to develop methodologies for quantifying environmental and socio-political factors involved with nuclear fuel cycle and finally to evaluate nuclear fuel cycle options with special emphasis given to the factors. Moreover, methodologies for developing practical radiological health risk assessment code system will be developed by which the assessment could be achieved for the recycling and reuse of scrap materials containing residual radioactive contamination. Selected scenarios are direct disposal, DUPIC(Direct use of PWR spent fuel in CANDU), and MOX recycle, land use, radiological effect, and non-radiological effect were chosen for environmental criteria and public acceptance and non-proliferation of nuclear material for socio-political ones. As a result of this study, potential scenarios to be chosen in Korea were selected and methodologies were developed to quantify the environmental and socio-political criteria. 24 refs., 27 tabs., 29 figs. (author)

  17. A study on the environmental friendliness of nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. J.; Lee, B. H.; Lee, S. Y.; Lim, C. Y.; Choi, Y. S.; Lee, Y. E.; Hong, D. S.; Cheong, J. H; Park, J. B.; Kim, K. K.; Cheong, H. Y; Song, M. C; Lee, H. J. [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of)

    1998-01-01

    The purpose of this study is to develop methodologies for quantifying environmental and socio-political factors involved with nuclear fuel cycle and finally to evaluate nuclear fuel cycle options with special emphasis given to the factors. Moreover, methodologies for developing practical radiological health risk assessment code system will be developed by which the assessment could be achieved for the recycling and reuse of scrap materials containing residual radioactive contamination. Selected scenarios are direct disposal, DUPIC(Direct use of PWR spent fuel in CANDU), and MOX recycle, land use, radiological effect, and non-radiological effect were chosen for environmental criteria and public acceptance and non-proliferation of nuclear material for socio-political ones. As a result of this study, potential scenarios to be chosen in Korea were selected and methodologies were developed to quantify the environmental and socio-political criteria. 24 refs., 27 tabs., 29 figs. (author)

  18. International Nuclear Fuel Cycle Fact Book

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.

    1992-05-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need exists costs for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book has been compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NMEA activities reports; and proceedings of conferences and workshops. The data listed typically do not reflect any single source but frequently represent a consolidation/combination of information.

  19. International Nuclear Fuel Cycle Fact Book

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I W; Mitchell, S J

    1990-01-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops, etc. The data listed do not reflect any one single source but frequently represent a consolidation/combination of information.

  20. International nuclear fuel cycle fact book

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.

    1988-01-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source or information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained has been obtained from nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops; and so forth. Sources do not agree completely with each other, and the data listed herein does not reflect any one single source but frequently is consolidation/combination of information. Lack of space as well as the intent and purpose of the Fact Book limit the given information to that pertaining to the Nuclear Fuel Cycle and to data considered of primary interest or most helpful to the majority of users.

  1. A comparison of mechanical algorithms of fuel performance code systems

    Energy Technology Data Exchange (ETDEWEB)

    Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C

    2003-11-01

    The goal of fuel rod performance evaluation is to identify the robustness of fuel rod with cladding material during fuel irradiation. Computer simulation of fuel rod performance becomes important to develop new nuclear systems. To construct the computing code system for fuel rod performance, we compared several algorithms of existing fuel rod performance code systems and summarized the details and tips as a preliminary work. Among several code systems, FRAPCON, FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. The computational algorithms related to mechanical interaction of the fuel rod are compared including methodologies and subroutines. This work will be utilized to develop the computing code system for dry process fuel rod performance.

  2. Sodium fast reactors with closed fuel cycle

    CERN Document Server

    Raj, Baldev; Vasudeva Rao, PR 0

    2015-01-01

    Sodium Fast Reactors with Closed Fuel Cycle delivers a detailed discussion of an important technology that is being harnessed for commercial energy production in many parts of the world. Presenting the state of the art of sodium-cooled fast reactors with closed fuel cycles, this book:Offers in-depth coverage of reactor physics, materials, design, safety analysis, validations, engineering, construction, and commissioning aspectsFeatures a special chapter on allied sciences to highlight advanced reactor core materials, specialized manufacturing technologies, chemical sensors, in-service inspecti

  3. FUEL CELL/MICRO-TURBINE COMBINED CYCLE

    Energy Technology Data Exchange (ETDEWEB)

    Larry J. Chaney; Mike R. Tharp; Tom W. Wolf; Tim A. Fuller; Joe J. Hartvigson

    1999-12-01

    A wide variety of conceptual design studies have been conducted that describe ultra-high efficiency fossil power plant cycles. The most promising of these ultra-high efficiency cycles incorporate high temperature fuel cells with a gas turbine. Combining fuel cells with a gas turbine increases overall cycle efficiency while reducing per kilowatt emissions. This study has demonstrated that the unique approach taken to combining a fuel cell and gas turbine has both technical and economic merit. The approach used in this study eliminates most of the gas turbine integration problems associated with hybrid fuel cell turbine systems. By using a micro-turbine, and a non-pressurized fuel cell the total system size (kW) and complexity has been reduced substantially from those presented in other studies, while maintaining over 70% efficiency. The reduced system size can be particularly attractive in the deregulated electrical generation/distribution environment where the market may not demand multi-megawatt central stations systems. The small size also opens up the niche markets to this high efficiency, low emission electrical generation option.

  4. Nuclear Fuel Cycle Options Catalog FY15 Improvements and Additions.

    Energy Technology Data Exchange (ETDEWEB)

    Price, Laura L. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Barela, Amanda Crystal [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Schetnan, Richard Reed [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Walkow, Walter M. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-11-01

    The United States Department of Energy, Office of Nuclear Energy, Fuel Cycle Technology Program sponsors nuclear fuel cycle research and development. As part of its Fuel Cycle Options campaign, the DOE has established the Nuclear Fuel Cycle Options Catalog. The catalog is intended for use by the Fuel Cycle Technologies Program in planning its research and development activities and disseminating information regarding nuclear energy to interested parties. The purpose of this report is to document the improvements and additions that have been made to the Nuclear Fuel Cycle Options Catalog in the 2015 fiscal year.

  5. San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    1999-09-01

    The isotopic composition of mixed-oxide fuel (fabricated with both uranium and plutonium isotope) discharged from reactors is of interest to the Fissile Material Disposition Program. The validation of depletion codes used to predict isotopic compositions of MOX fuel, similar to studies concerning uranium-only fueled reactors, thus, is very important. The EEI-Westinghouse Plutonium Recycle Demonstration Program was conducted to examine the use of MOX fuel in the San Onofre PWR, Unit I, during cycles 2 and 3. The data usually required as input to depletion codes, either one-dimensional or lattice codes, were taken from various sources and compiled into this report. Where data were either lacking or determined inadequate, the appropriate data were supplied from other references. The scope of the reactor operations and design data, in addition to the isotopic analyses, were considered to be of sufficient quality for depletion code validation.

  6. International nuclear fuel cycle fact book. Revision 6

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.; Jeffs, A.G.

    1986-01-01

    The International Fuel Cycle Fact Book has been compiled in an effort to provide (1) an overview of worldwide nuclear power and fuel cycle programs and (2) current data concerning fuel cycle and waste management facilities, R and D programs and key personnel. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2.

  7. Non-judgemental Dynamic Fuel Cycle Benchmarking

    CERN Document Server

    Scopatz, Anthony Michael

    2015-01-01

    This paper presents a new fuel cycle benchmarking analysis methodology by coupling Gaussian process regression, a popular technique in Machine Learning, to dynamic time warping, a mechanism widely used in speech recognition. Together they generate figures-of-merit that are applicable to any time series metric that a benchmark may study. The figures-of-merit account for uncertainty in the metric itself, utilize information across the whole time domain, and do not require that the simulators use a common time grid. Here, a distance measure is defined that can be used to compare the performance of each simulator for a given metric. Additionally, a contribution measure is derived from the distance measure that can be used to rank order the importance of fuel cycle metrics. Lastly, this paper warns against using standard signal processing techniques for error reduction. This is because it is found that error reduction is better handled by the Gaussian process regression itself.

  8. Fuel Cycle Technologies 2014 Achievement Report

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Bonnie C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-01-01

    The Fuel Cycle Technologies (FCT) program supports the Department of Energy’s (DOE’s) mission to: “Enhance U.S. security and economic growth through transformative science, technology innovation, and market solutions to meet our energy, nuclear security, and environmental challenges.” Goal 1 of DOE’s Strategic Plan is to innovate energy technologies that enhance U.S. economic growth and job creation, energy security, and environmental quality. FCT does this by investing in advanced technologies that could transform the nuclear fuel cycle in the decades to come. Goal 2 of DOE’s Strategic Plan is to strengthen national security by strengthening key science, technology, and engineering capabilities. FCT does this by working closely with the National Nuclear Security Administration and the U.S Department of State to develop advanced technologies that support the Nation’s nuclear nonproliferation goals.

  9. Current Comparison of Advanced Nuclear Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    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

  10. HTR Spherical Super Lattice Model for Equilibrium Fuel Cycle Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gray S. Cahng

    2005-09-01

    Advanced High Temperature gas-cooled Reactors (HTR) currently being developed (GFR, VHTR - Very High Temperature gas-cooled Reactor, PBMR, and GT-MHR) are able to achieve a simplification of safety through reliance on innovative features and passive systems. One of the innovative features in these HTRs is reliance on ceramic-coated fuel particles to retain the fission products even under extreme accident conditions. The effect of the random fuel kernel distribution in the fuel pebble / block is addressed through the use of the Dancoff correction factor in the resonance treatment. In addition, the Dancoff correction factor is a function of burnup and fuel kernel packing factor, which requires that the Dancoff correction factor be updated during Equilibrium Fuel Cycle (EqFC) analysis. Although HTR fuel is rather homogeneously dispersed in the fuel graphite matrix, the heterogeneity effects in between fuel kernels and pebbles cannot be ignored. The double-heterogeneous lattice model recently developed at the Idaho National Engineering and Environmental Laboratory (INEEL) contains tens of thousands of cubic fuel kernel cells, which makes it very difficult to deplete the fuel, kernel by kernel (KbK), for the EqFC analysis. In addition, it is not possible to preserve the cubic size and packing factor in a spherical fuel pebble. To avoid these difficulties, a newly developed and validated HTR pebble-bed Kernel-by-Kernel spherical (KbK-sph) model, has been developed and verified in this study. The objective of this research is to introduce the KbK-sph model and super whole Pebble lattice model (PLM). The verified double-heterogeneous KbK-sph and pebble homogeneous lattice model (HLM) are used for the fuel burnup chracteristics analysis and important safety parameters validation. This study summarizes and compares the KbK-sph and HLM burnup analyzed results. Finally, we discus the Monte-Carlo coupling with a fuel depletion and buildup code - Origen-2 as a fuel burnup

  11. Closed Fuel Cycle Waste Treatment Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Vienna, J. D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Collins, E. D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Crum, J. V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ebert, W. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Frank, S. M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Garn, T. G. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gombert, D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jones, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jubin, R. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Maio, V. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Matyas, J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Nenoff, T. M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Riley, B. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sevigny, G. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Soelberg, N. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strachan, D. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Thallapally, P. K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westsik, J. H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-02-01

    This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form

  12. A fuel management study and cycle nuclear design for PW reactors

    Energy Technology Data Exchange (ETDEWEB)

    Minguez, E.; Ahnert, C.; Aragones, J. M.; Corella, M. R.

    1975-07-01

    A reference reactor was chosen to do a general study involving Fuel Management Evaluations of several cycles, and Design Calculations of cycles already performed, according to a calculation scheme set up in the Reactor Technology Division of the J.E.N., using some computer codes acquired to foreign sources and other ones developed in the J.E.N. (Author) 5 refs.

  13. The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

    Science.gov (United States)

    Saldikov, I. S.; Ternovykh, M. Yu; Fomichenko, P. A.; Gerasimov, A. S.

    2017-01-01

    The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

  14. Survey of nuclear fuel cycle economics: 1970--1985

    Energy Technology Data Exchange (ETDEWEB)

    Prince, B. E.; Peerenboom, J. P.; Delene, J. G.

    1977-03-01

    This report is intended to provide a coherent view of the diversity of factors that may affect nuclear fuel cycle economics through about 1985. The nuclear fuel cycle was surveyed as to past trends, current problems, and future considerations. Unit costs were projected for each step in the fuel cycle. Nuclear fuel accounting procedures were reviewed; methods of calculating fuel costs were examined; and application was made to Light Water Reactors (LWR) over the next decade. A method conforming to Federal Power Commission accounting procedures and used by utilities to account for backend fuel-cycle costs was described which assigns a zero net salvage value to discharged fuel. LWR fuel cycle costs of from 4 to 6 mills/kWhr (1976 dollars) were estimated for 1985. These are expected to reach 6 to 9 mills/kWr if the effect of inflation is included.

  15. Compatibility analysis of DUPIC fuel (part5) - DUPIC fuel cycle economics analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Choi, Hang Bok; Yang, Myung Seung

    2000-08-01

    This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the one-batch equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.55-6.72 mills/kWh for proposed DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.04-0.28 mills/kWh. Considering the uncertainty (0.45-0.51 mills/kWh) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by -20% and reduce the spent fuel arising by -65%, compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle possesses a strong advantage over the once-through fuel cycle from the viewpoint of the environmental effect.

  16. Uranium to Electricity: The Chemistry of the Nuclear Fuel Cycle

    Science.gov (United States)

    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…

  17. Predictive Bias and Sensitivity in NRC Fuel Performance Codes

    Energy Technology Data Exchange (ETDEWEB)

    Geelhood, Kenneth J.; Luscher, Walter G.; Senor, David J.; Cunningham, Mitchel E.; Lanning, Donald D.; Adkins, Harold E.

    2009-10-01

    The latest versions of the fuel performance codes, FRAPCON-3 and FRAPTRAN were examined to determine if the codes are intrinsically conservative. Each individual model and type of code prediction was examined and compared to the data that was used to develop the model. In addition, a brief literature search was performed to determine if more recent data have become available since the original model development for model comparison.

  18. New Mechanical Model for the Transmutation Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Gregory K. Miller

    2008-04-01

    A new mechanical model has been developed for implementation into the TRU fuel performance code. The new model differs from the existing FRAPCON 3 model, which it is intended to replace, in that it will include structural deformations (elasticity, plasticity, and creep) of the fuel. Also, the plasticity algorithm is based on the “plastic strain–total strain” approach, which should allow for more rapid and assured convergence. The model treats three situations relative to interaction between the fuel and cladding: (1) an open gap between the fuel and cladding, such that there is no contact, (2) contact between the fuel and cladding where the contact pressure is below a threshold value, such that axial slippage occurs at the interface, and (3) contact between the fuel and cladding where the contact pressure is above a threshold value, such that axial slippage is prevented at the interface. The first stage of development of the model included only the fuel. In this stage, results obtained from the model were compared with those obtained from finite element analysis using ABAQUS on a problem involving elastic, plastic, and thermal strains. Results from the two analyses showed essentially exact agreement through both loading and unloading of the fuel. After the cladding and fuel/clad contact were added, the model demonstrated expected behavior through all potential phases of fuel/clad interaction, and convergence was achieved without difficulty in all plastic analysis performed. The code is currently in stand alone form. Prior to implementation into the TRU fuel performance code, creep strains will have to be added to the model. The model will also have to be verified against an ABAQUS analysis that involves contact between the fuel and cladding.

  19. Fuel burnup analysis for Thai research reactor by using MCNPX computer code

    Science.gov (United States)

    Sangkaew, S.; Angwongtrakool, T.; Srimok, B.

    2017-06-01

    This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.

  20. User's guide for the REBUS-3 fuel cycle analysis capability

    Energy Technology Data Exchange (ETDEWEB)

    Toppel, B.J.

    1983-03-01

    REBUS-3 is a system of programs designed for the fuel-cycle analysis of fast reactors. This new capability is an extension and refinement of the REBUS-3 code system and complies with the standard code practices and interface dataset specifications of the Committee on Computer Code Coordination (CCCC). The new code is hence divorced from the earlier ARC System. In addition, the coding has been designed to enhance code exportability. Major new capabilities not available in the REBUS-2 code system include a search on burn cycle time to achieve a specified value for the multiplication constant at the end of the burn step; a general non-repetitive fuel-management capability including temporary out-of-core fuel storage, loading of fresh fuel, and subsequent retrieval and reloading of fuel; significantly expanded user input checking; expanded output edits; provision of prestored burnup chains to simplify user input; option of fixed-or free-field BCD input formats; and, choice of finite difference, nodal or spatial flux-synthesis neutronics in one-, two-, or three-dimensions.

  1. Life cycle analysis of transportation fuel pathways

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-02-24

    The purpose of this work is to improve the understanding of the concept of life cycle analysis (LCA) of transportation fuels and some of its pertinent issues among non-technical people, senior managers, and policy makers. This work should provide some guidance to nations considering LCA-based policies and to people who are affected by existing policies or those being developed. While the concept of employing LCA to evaluate fuel options is simple and straightforward, the act of putting the concept into practice is complex and fraught with issues. Policy makers need to understand the limitations inherent in carrying out LCA work for transportation fuel systems. For many systems, even those that have been employed for a 100 years, there is a lack of sound data on the performance of those systems. Comparisons between systems should ideally be made using the same tool, so that differences caused by system boundaries, allocation processes, and temporal issues can be minimized (although probably not eliminated). Comparing the results for fuel pathway 1 from tool A to those of fuel system 2 from tool B introduces significant uncertainty into the results. There is also the question of the scale of system changes. LCA will give more reliable estimates when it is used to examine small changes in transportation fuel pathways than when used to estimate large scale changes that replace current pathways with completely new pathways. Some LCA tools have been developed recently primarily for regulatory purposes. These tools may deviate from ISO principles in order to facilitate simplicity and ease of use. In a regulatory environment, simplicity and ease of use are worthy objectives and in most cases there is nothing inherently wrong with this approach, particularly for assessing relative performance. However, the results of these tools should not be confused with, or compared to, the results that are obtained from a more complex and rigorous ISO compliant LCA. It should be

  2. Study of Thorium Fuel Cycles for Light Water Reactor VBER-150

    Directory of Open Access Journals (Sweden)

    Daniel Evelio Milian Lorenzo

    2013-01-01

    Full Text Available The main objective of this paper is to examine the use of thorium-based fuel cycle for the transportable reactors or transportable nuclear power plants (TNPP VBER-150 concept, in particular the neutronic behavior. The thorium-based fuel cycles included Th232+Pu239, Th232+U233, and Th232+U and the standard design fuel UOX. Parameters related to the neutronic behavior such as burnup, nuclear fuel breeding, MA stockpile, and Pu isotopes production (among others were used to compare the fuel cycles. The Pu transmutation rate and accumulation of Pu with MA in the spent fuel were compared mutually and with an UOX open cycle. The Th232+U233 fuel cycle proved to be the best cycle for minimizing the production of Pu and MA. The neutronic calculations have been performed with the well-known MCNPX computational code, which was verified for this type of fuel performing calculation of the IAEA benchmark announced by IAEA-TECDOC-1349.

  3. Fuel cycle assessment: A compendium of models, methodologies, and approaches

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-01

    The purpose of this document is to profile analytical tools and methods which could be used in a total fuel cycle analysis. The information in this document provides a significant step towards: (1) Characterizing the stages of the fuel cycle. (2) Identifying relevant impacts which can feasibly be evaluated quantitatively or qualitatively. (3) Identifying and reviewing other activities that have been conducted to perform a fuel cycle assessment or some component thereof. (4) Reviewing the successes/deficiencies and opportunities/constraints of previous activities. (5) Identifying methods and modeling techniques/tools that are available, tested and could be used for a fuel cycle assessment.

  4. ATR PDQ and MCWO Fuel Burnup Analysis Codes Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    G.S. Chang; P. A. Roth; M. A. Lillo

    2009-11-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is being studied to determine the feasibility of converting it from the highly enriched Uranium (HEU) fuel that is currently uses to low enriched Uranium (LEU) fuel. In order to achieve this goal, it would be best to qualify some different computational methods than those that have been used at ATR for the past 40 years. This paper discusses two methods of calculating the burnup of ATR fuel elements. The existing method, that uses the PDQ code, is compared to a modern method that uses A General Monte Carlo N-Particle Transport Code (MCNP) combined with the Origen2.2 code. This modern method, MCNP with ORIGEN2.2 (MCWO), is found to give excellent agreement with the existing method (PDQ). Both of MCWO and PDQ are also in a very good agreement to the 235U burnup data generated by an analytical method.

  5. Results of studies on application of CCMHD to advanced fossil fuel power plant cycles

    Energy Technology Data Exchange (ETDEWEB)

    Foote, J.P.; Wu, Y.C.L.S.; Lineberry, J.T.

    1998-07-01

    A study was conducted to assess the potential for application of a Closed Cycle MHD disk generator (CCMHD) in advanced fossil fuel power generation systems. Cycle analyses were conducted for a variety of candidate power cycles, including simple cycle CCMHD (MHD); a cycle combining CCMHD and gas turbines (MHD/GT); and a triple combined cycle including CCMHD, gas turbines, and steam turbines (MHD/GT/ST). The above cycles were previously considered in cycle studies reported by Japanese researchers. Also considered was a CCMHD cycle incorporating thermochemical heat recovery through reforming of the fuel stream (MHD/REF), which is the first consideration of this approach. A gas turbine/steam turbine combined cycle (GT/ST) was also analyzed for baseline comparison. The only fuel considered in the study was CH4. Component heat and pressure losses were neglected, and the potential for NOx emission due to high combustion temperatures was not considered. Likewise, engineering limitations for cycle components, particularly the high temperature argon heater, were not considered. This approach was adopted to simplify the analysis for preliminary screening of candidate cycles. Cycle calculations were performed using in-house code. Ideal gas thermodynamic properties were calculated using the NASA SP- 273 data base, and thermodynamic properties for steam were calculated using the computerized ASME Steam Tables. High temperature equilibrium compositions for combustion gas were calculated using tabulated values of the equilibrium constants for the important reactions.

  6. Software Requirements Specification Verifiable Fuel Cycle Simulation (VISION) Model

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Environmental Emissions From Energy Technology Systems: The Total Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    San Martin, Robert L.

    1989-04-01

    This is a summary report that compares emissions during the entire project life cycle for a number of fossil-fueled and renewable electric power systems, including geothermal steam (probably modeled after The Geysers). The life cycle is broken into Fuel Extraction, Construction, and Operation. The only emission covered is carbon dioxide. (DJE 2005)

  8. Preliminary analysis of alternative fuel cycles for proliferation evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Steindler, M. J.; Ripfel, H. C.F.; Rainey, R. H.

    1977-01-01

    The ERDA Division of Nuclear Research and Applications proposed 67 nuclear fuel cycles for assessment as to their nonproliferation potential. The object of the assessment was to determine which fuel cycles pose inherently low risk for nuclear weapon proliferation while retaining the major benefits of nuclear energy. This report is a preliminary analysis of these fuel cycles to develop the fuel-recycle data that will complement reactor data, environmental data, and political considerations, which must be included in the overall evaluation. This report presents the preliminary evaluations from ANL, HEDL, ORNL, and SRL and is the basis for a continuing in-depth study. (DLC)

  9. Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

    Science.gov (United States)

    Tikhomirov, Georgy; Ternovykh, Mikhail; Saldikov, Ivan; Fomichenko, Peter; Gerasimov, Alexander

    2017-09-01

    The strategy of the development of nuclear power in Russia provides for use of fast power reactors in closed nuclear fuel cycle. The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of energy. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. The closed nuclear fuel cycle concept of the PRORYV assumes self-supplied mode of operation with fuel regeneration by neutron capture reaction in non-enriched uranium, which is used as a raw material. Operating modes of reactors and its characteristics should be chosen so as to provide the self-sufficient mode by using of fissile isotopes while refueling by depleted uranium and to support this state during the entire period of reactor operation. Thus, the actual issue is modeling fuel handling processes. To solve these problems, the code REPRORYV (Recycle for PRORYV) has been developed. It simulates nuclide streams in non-reactor stages of the closed fuel cycle. At the same time various verified codes can be used to evaluate in-core characteristics of a reactor. By using this approach various options for nuclide streams and assess the impact of different plutonium content in the fuel, fuel processing conditions, losses during fuel processing, as well as the impact of initial uncertainties on neutron-physical characteristics of reactor are considered in this study.

  10. Licos, a fuel performance code for innovative fuel elements or experimental devices design

    Energy Technology Data Exchange (ETDEWEB)

    Helfer, Thomas, E-mail: thomas.helfer@cea.fr; Bejaoui, Syriac, E-mail: syriac.bejaoui@cea.fr; Michel, Bruno, E-mail: bruno.michel@cea.fr

    2015-12-01

    Highlights: • The Licos fuel performance code is introduced. • Advanced features, such as dependency algorithm and kriging are described. • First results on three dimensional modelling of the SFR fuel pin are given. • Application to the DIAMINO design computations is discussed. - Abstract: This paper provides an overview of the Licos fuel performance code which has been developed for several years within the platform pleiades, co-developed by the French Alternative Energies and Atomic Energy Commission (CEA) and its industrial partners Électricité de France (EDF) and AREVA. CEA engineers have been using Licos to back multidimensional thermo-mechanical studies on innovative fuel elements design and experimental device pre-and post-irradiation computations. Studies made with Licos thus encompass a wide range of situations, including most nuclear systems used or studied in France in recent years (PWR, SFR or GFR), normal and off-normal operating conditions, and a large selection of materials (either for fuel, absorber, coolant and cladding). The aim of this paper is to give some insights about some innovative features in the design of Licos (dependency management, kriging, mfront, etc.). We also present two studies that demonstrate the flexibility of this code. The first one shows how Licos can be combined with the Germinal monodimensional fuel performance code to demonstrate the interest of a three dimensional modelling of the fuel relocation phenomenon in the Sodium Fast Reactor fuel pin. The second one describes how Licos was used to model the DIAMINO experiment.

  11. Sustainable thorium nuclear fuel cycles: A comparison of intermediate and fast neutron spectrum systems

    Energy Technology Data Exchange (ETDEWEB)

    Brown, N.R., E-mail: nbrown@bnl.gov [Brookhaven National Laboratory, Upton, NY (United States); Powers, J.J. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Feng, B.; Heidet, F.; Stauff, N.E.; Zhang, G. [Argonne National Laboratory, Argonne, IL (United States); Todosow, M. [Brookhaven National Laboratory, Upton, NY (United States); Worrall, A.; Gehin, J.C. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Kim, T.K.; Taiwo, T.A. [Argonne National Laboratory, Argonne, IL (United States)

    2015-08-15

    Highlights: • Comparison of intermediate and fast spectrum thorium-fueled reactors. • Variety of reactor technology options enables self-sustaining thorium fuel cycles. • Fuel cycle analyses indicate similar performance for fast and intermediate systems. • Reproduction factor plays a significant role in breeding and burn-up performance. - Abstract: This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10{sup 5} eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight lattice heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this self-sustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems.

  12. 77 FR 65729 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

    Science.gov (United States)

    2012-10-30

    ... COMMISSION Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC... inspections of the Louisiana Energy Services (LES), LLC, National Enrichment Facility in Eunice, New Mexico... Title 10 of the Code of Federal Regulations (10 CFR) 70.32 (k) and section 193(c) of the Atomic...

  13. Overview of the BISON Multidimensional Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    R. L. Williamson; J. D. Hales; S. R. Novascone; B. W. Spencer; D. M. Perez; G. Pastore; R. C. Martineau

    2013-10-01

    BISON is a modern multidimensional multiphysics finite-element based nuclear fuel performance code that has been under development at the Idaho National Laboratory (USA) since 2009. A brief background is provided on the code’s computational framework (MOOSE), governing equations, and material and behavioral models. Ongoing code verification and validation work is outlined, and comparative results are provided for select validation cases. Recent applications are discussed, including specific description of two applications where 3D treatment is important. A summary of future code development and validation activities is given. Numerous references to published work are provided where interested readers can find more complete information.

  14. FGR Evaluation Code for CANDU Fuel under an Accidents: REDOU

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jung, Jong Yeob

    2006-12-15

    When an end fitting failure is occurred, the sheath is broken and the fission gases are released promptly from the gap between the pellet and sheath and also released continuously from the inside of the pellet due to the oxidation of the pellet. Thus, the accurate calculation of the fission gas release from the gap and from the inside of the pellet is an essential to the accident analysis of the fuel behavior and to prepare the safety strategy. In this report, among the performance analysis or the transient behavior prediction computer codes, the REDOU code which is the fission gas calculating code for the postulated accident scenario such as an end fitting failure is introduced. Then, the user manual for REDOU code is provided so that it can be the guidance to the potential users of the code and save the time and economic loss by reducing the trial and error.

  15. Feasibility of Thorium Fuel Cycles in a Very High Temperature Pebble-Bed Hybrid System

    Directory of Open Access Journals (Sweden)

    L.P. Rodriguez

    2015-08-01

    Full Text Available Nuclear energy presents key challenges to be successful as a sustainable energy source. Currently, the viability of the use thorium-based fuel cycles in an innovative nuclear energy generation system is being investigated in order to solve these key challenges. In this work, the feasibility of three thorium-based fuel cycles (232Th-233U, 232Th-239Pu, and 232Th-U in a hybrid system formed by a Very High Temperature Pebble-Bed Reactor (VHTR and two Pebble-Bed Accelerator Driven Systems (ADSs was evaluated using parameters related to the neutronic behavior such as nuclear fuel breeding, minor actinide stockpile, the energetic contribution of each fissile isotope, and the radiotoxicity of the long lived wastes. These parameters were used to compare the fuel cycles using the well-known MCNPX ver. 2.6e computational code. The results obtained confirm that the 232Th-233U fuel cycle is the best cycle for minimizing the production of plutonium isotopes and minor actinides. Moreover, the inclusion of the second stage in the ADSs demonstrated the possibility of extending the burnup cycle duration and reducing the radiotoxicity of the discharged fuel from the VHTR.

  16. Nuclear fuel cycle facility accident analysis handbook

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The purpose of this Handbook is to provide guidance on how to calculate the characteristics of releases of radioactive materials and/or hazardous chemicals from nonreactor nuclear facilities. In addition, the Handbook provides guidance on how to calculate the consequences of those releases. There are four major chapters: Hazard Evaluation and Scenario Development; Source Term Determination; Transport Within Containment/Confinement; and Atmospheric Dispersion and Consequences Modeling. These chapters are supported by Appendices, including: a summary of chemical and nuclear information that contains descriptions of various fuel cycle facilities; details on how to calculate the characteristics of source terms for releases of hazardous chemicals; a comparison of NRC, EPA, and OSHA programs that address chemical safety; a summary of the performance of HEPA and other filters; and a discussion of uncertainties. Several sample problems are presented: a free-fall spill of powder, an explosion with radioactive release; a fire with radioactive release; filter failure; hydrogen fluoride release from a tankcar; a uranium hexafluoride cylinder rupture; a liquid spill in a vitrification plant; and a criticality incident. Finally, this Handbook includes a computer model, LPF No.1B, that is intended for use in calculating Leak Path Factors. A list of contributors to the Handbook is presented in Chapter 6. 39 figs., 35 tabs.

  17. Filling Knowledge Gaps with Five Fuel Cycle Studies

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Jess Gehin; William Halsey; Temitope Taiwo

    2010-11-01

    During FY 2010, five studies were conducted of technology families’ applicability to various fuel cycle strategies to fill in knowledge gaps in option space and to better understand trends and patterns. Here, a “technology family” is considered to be defined by a type of reactor and by selection of which actinides provide fuel. This report summarizes the higher-level findings; the detailed analyses and results are documented in five individual reports, as follows: • Advanced once through with uranium fuel in fast reactors (SFR), • Advanced once through (uranium fuel) or single recycle (TRU fuel) in high temperature gas cooled reactors (HTGR), • Sustained recycle with Th/U-233 in light water reactors (LWRs), • Sustained recycle with Th/U-233 in molten salt reactors (MSR), and • Several fuel cycle missions with Fusion-Fission Hybrid (FFH). Each study examined how the designated technology family could serve one or more designated fuel cycle missions, filling in gaps in overall option space. Each study contains one or more illustrative cases that show how the technology family could be used to meet a fuel cycle mission, as well as broader information on the technology family such as other potential fuel cycle missions for which insufficient information was available to include with an illustrative case. None of the illustrative cases can be considered as a reference, baseline, or nominal set of parameters for judging performance; the assessments were designed to assess areas of option space and were not meant to be optimized. There is no implication that any of the cases or technology families are necessarily the best way to meet a given fuel cycle mission. The studies provide five examples of 1-year fuel cycle assessments of technology families. There is reasonable coverage in the five studies of the performance areas of waste management and uranium utilization. The coverage of economics, safety, and proliferation resistance and physical protection in

  18. Analysis of Proliferation Resistance of Nuclear Fuel Cycle Systems

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Hong Lae; Ko, Won Il; Kim, Ho Dong

    2009-11-15

    Proliferation resistance (PR) has been evaluated for the five nuclear fuel cycle systems, potentially deployable in Korea in the future, using the fourteen proliferation resistance attributes suggested in the TOPS report. Unidimensional Utility Theory (UUT) was used in the calculation of utility value for each of the fourteen proliferation resistance attributes, and Multi-Attribute Utility Theory (MAUT), a decision tool with multiple objectives, was used in the evaluation of the proliferation resistance of each nuclear fuel cycle system. Analytic Hierarchy Process (AHP) and Expert Elicitation (EE) were utilized in the derivation of weighting factors for the fourteen proliferation resistance attributes. Among the five nuclear fuel cycle systems evaluated, the once-through fuel cycle system showed the highest level of proliferation resistance, and Pyroprocessing-SFR fuel cycle system showed the similar level of proliferation resistance with the DUPIC fuel cycle system, which has two time higher level of proliferation resistance compared to that of the thermal MOX fuel cycle system. Sensitivity analysis was also carried out to make up for the uncertainty associated with the derivation of weighting factors for the fourteen proliferation resistance attributes.

  19. Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations

    Energy Technology Data Exchange (ETDEWEB)

    Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    A burn-up calculation of VVER's cores by Monte-Carlo code is complex process and requires large computational costs. This fact makes Monte-Carlo codes usage complicated for project and operating calculations. Previously prepared isotopic compositions are proposed to use for the Monte-Carlo code (MCU) calculations of different states of VVER's core with burnt fuel. Isotopic compositions are proposed to calculate by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by engineering codes (TVS-M, PERMAK-A). The multiplication factors and power distributions of FA and VVER with infinite height are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The MCU calculation data were compared with the data which were obtained by engineering codes.

  20. Technology Insights and Perspectives for Nuclear Fuel Cycle Concepts

    Energy Technology Data Exchange (ETDEWEB)

    S. Bays; S. Piet; N. Soelberg; M. Lineberry; B. Dixon

    2010-09-01

    The following report provides a rich resource of information for exploring fuel cycle characteristics. The most noteworthy trends can be traced back to the utilization efficiency of natural uranium resources. By definition, complete uranium utilization occurs only when all of the natural uranium resource can be introduced into the nuclear reactor long enough for all of it to undergo fission. Achieving near complete uranium utilization requires technologies that can achieve full recycle or at least nearly full recycle of the initial natural uranium consumed from the Earth. Greater than 99% of all natural uranium is fertile, and thus is not conducive to fission. This fact requires the fuel cycle to convert large quantities of non-fissile material into fissile transuranics. Step increases in waste benefits are closely related to the step increase in uranium utilization going from non-breeding fuel cycles to breeding fuel cycles. The amount of mass requiring a disposal path is tightly coupled to the quantity of actinides in the waste stream. Complete uranium utilization by definition means that zero (practically, near zero) actinide mass is present in the waste stream. Therefore, fuel cycles with complete (uranium and transuranic) recycle discharge predominately fission products with some actinide process losses. Fuel cycles without complete recycle discharge a much more massive waste stream because only a fraction of the initial actinide mass is burned prior to disposal. In a nuclear growth scenario, the relevant acceptable frequency for core damage events in nuclear reactors is inversely proportional to the number of reactors deployed in a fuel cycle. For ten times the reactors in a fleet, it should be expected that the fleet-average core damage frequency be decreased by a factor of ten. The relevant proliferation resistance of a fuel cycle system is enhanced with: decreasing reliance on domestic fuel cycle services, decreasing adaptability for technology misuse

  1. Modelling of LOCA Tests with the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Richard L [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory

    2016-05-01

    BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

  2. Current Capabilities of the Fuel Performance Modeling Code PARFUME

    Energy Technology Data Exchange (ETDEWEB)

    G. K. Miller; D. A. Petti; J. T. Maki; D. L. Knudson

    2004-09-01

    The success of gas reactors depends upon the safety and quality of the coated particle fuel. A fuel performance modeling code (called PARFUME), which simulates the mechanical and physico-chemical behavior of fuel particles during irradiation, is under development at the Idaho National Engineering and Environmental Laboratory. Among current capabilities in the code are: 1) various options for calculating CO production and fission product gas release, 2) a thermal model that calculates a time-dependent temperature profile through a pebble bed sphere or a prismatic block core, as well as through the layers of each analyzed particle, 3) simulation of multi-dimensional particle behavior associated with cracking in the IPyC layer, partial debonding of the IPyC from the SiC, particle asphericity, kernel migration, and thinning of the SiC caused by interaction of fission products with the SiC, 4) two independent methods for determining particle failure probabilities, 5) a model for calculating release-to-birth (R/B) ratios of gaseous fission products, that accounts for particle failures and uranium contamination in the fuel matrix, and 6) the evaluation of an accident condition, where a particle experiences a sudden change in temperature following a period of normal irradiation. This paper presents an overview of the code.

  3. EHR's effect on the revenue cycle management Coding function.

    Science.gov (United States)

    Giannangelo, Kathy; Fenton, Susan

    2008-01-01

    Without administrative terminologies there is no revenue to manage. The use of healthcare IT to capture the codes for administrative and financial support functions will impact the revenue cycle and the management of it. This is presumed to occur because clinical data coded at the point of care becomes the source for claims data. Thus, as electronic health record system applications utilizing terminologies are implemented, healthcare providers need to systematically consider the effect on the coding function and management of the revenue cycle. A key factor is the sequence of events changes, i.e., instead of a health information management professional selecting billing codes at the conclusion of an encounter based on the review of the record, clinical data generates the claims data via mapping. Efficiencies and management challenges result.

  4. AB 1007 Full Fuel Cycle Analysis (FFCA) Peer Review

    Energy Technology Data Exchange (ETDEWEB)

    Rice, D; Armstrong, D; Campbell, C; Lamont, A; Gallegos, G; Stewart, J; Upadhye, R

    2007-01-19

    LLNL is a participant of California's Advanced Energy Pathways (AEP) team funded by DOE (NETL). At the AEP technical review meeting on November 9, 2006. The AB 1007 FFCA team (Appendix A) requested LLNL participate in a peer review of the FFCA reports. The primary contact at the CEC was McKinley Addy. The following reports/presentations were received by LLNL: (1) Full Fuel Cycle Energy and Emissions Assumptions dated September 2006, TIAX; (2) Full Fuel cycle Assessment-Well to Tank Energy Inputs, Emissions, and Water Impacts dated December 2006, TIAX; and (3) Full Fuel Cycle Analysis Assessment dated October 12, 2006, TIAX.

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

    Science.gov (United States)

    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

  6. Intercode Advanced Fuels and Cladding Comparison Using BISON, FRAPCON, and FEMAXI Fuel Performance Codes

    Science.gov (United States)

    Rice, Aaren

    As part of the Department of Energy's Accident Tolerant Fuels (ATF) campaign, new cladding designs and fuel types are being studied in order to help make nuclear energy a safer and more affordable source for power. This study focuses on the implementation and analysis of the SiC cladding and UN, UC, and U3Si2 fuels into three specific nuclear fuel performance codes: BISON, FRAPCON, and FEMAXI. These fuels boast a higher thermal conductivity and uranium density than traditional UO2 fuel which could help lead to longer times in a reactor environment. The SiC cladding has been studied for its reduced production of hydrogen gas during an accident scenario, however the SiC cladding is a known brittle and unyielding material that may fracture during PCMI (Pellet Cladding Mechanical Interaction). This work focuses on steady-state operation with advanced fuel and cladding combinations. By implementing and performing analysis work with these materials, it is possible to better understand some of the mechanical interactions that could be seen as limiting factors. In addition to the analysis of the materials themselves, a further analysis is done on the effects of using a fuel creep model in combination with the SiC cladding. While fuel creep is commonly ignored in the traditional UO2 fuel and Zircaloy cladding systems, fuel creep can be a significant factor in PCMI with SiC.

  7. Information sets as permutation cycles for quadratic residue codes

    Directory of Open Access Journals (Sweden)

    Richard A. Jenson

    1982-01-01

    Full Text Available The two cases p=7 and p=23 are the only known cases where the automorphism group of the [p+1,   (p+1/2] extended binary quadratic residue code, O(p, properly contains PSL(2,p. These codes have some of their information sets represented as permutation cycles from Aut(Q(p. Analysis proves that all information sets of Q(7 are so represented but those of Q(23 are not.

  8. Nonproliferation and safeguards aspects of the DUPIC fuel cycle concept

    Energy Technology Data Exchange (ETDEWEB)

    Persiani, P. K. [Argonne National Lab., IL (United States)

    1997-07-01

    The purpose of the study is to comment on the proliferation characteristic profiles of some of the proposed fuel cycle alternatives to help ensure that nonproliferation concerns are introduced into the early stages of a fuel cycle concept development program, and to perhaps aid in the more effective implementation of the international nonproliferation regime initiative and safeguards systems. Alternative recycle concepts proposed by several countries involve the recycle of spent fuel without the separation of plutonium from uranium and fission products. The concepts are alternatives to either the direct long-term storage deposition of or the purex reprocessing of the spent fuels. The alternate fuel cycle concepts reviewed include: the dry-recycle processes such as the direct use of reconfigured PWR spent fuel assemblies into CANDU reactors(DUPIC); low-decontamination, single-cycle co-extraction of fast reactor fuels in a wet-purex type of reprocessing; and on a limited scale the thorium-uranium fuel cycle. The nonproliferation advantages usually associated with the above non-separation processes are: the highly radioactive spent fuel presents a barrier to the physical diversion of the nuclear material; avoid the need to dissolve and chemically separate the plutonium from the uranium and fission products; and that the spent fuel isotopic quality of the plutonium vector is further degraded. Although the radiation levels and the need for reprocessing may be perceived as barriers to the terrorist or the subnational level of safeguards, the international level of nonproliferation concerns is addressed primarily by material accountancy and verification activities. On the international level of nonproliferation concerns, the non-separation fuel cycle concepts involved have to be evaluated on the bases of the impact the processes may have on nuclear materials accountancy. (author).

  9. Direction of Improvement for Licensing Advanced Fuel Cycle Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Il; Yook, Dae Sik; Jeong, Seung Yeong; Jeong, Chan Woo [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    The final spent fuel management method has yet to be established, and the South Korean government is expected to decide on the final spent fuel management method under a national consensus. In particular, two methods of spent fuel management are under consideration: Direct disposal in base rock several hundred meters underground and recycling. The present study reviewed the direction of improvement of the regulatory system that can be applied when an advanced fuel cycle for recycling spent fuel is adopted as the final management method. For recycling to be adopted as the domestic final spent fuel management method, there remains the task of having to overcome the stumbling blocks of a national consensus and the Agreement for Cooperation between the Government of the Republic of Korea and the Government of the United States of America concerning the Civil Use of Atomic Energy (Korea-US Atomic Energy Agreement). To resolve this and to construct and operate advanced fuel cycle facilities, it is necessary to establish an applicable legal system, which the present study reviewed. The results of the present study are expected to be used as the basic data in improving the legal system after the realization of advanced fuel cycles in the future. In addition, research on the development of technical standards and safety requirements for advanced fuel cycle facilities will continue to be necessary.

  10. Model of U3Si2 Fuel System using BISON Fuel Code

    Energy Technology Data Exchange (ETDEWEB)

    K. E. Metzger; T. W. Knight; R. L. Williamson

    2014-04-01

    This research considers the proposed advanced fuel system: U3Si2 combined with an advanced cladding. U3Si2 has a number of advantageous thermophysical properties, which motivate its use as an accident tolerant fuel. This preliminary model evaluates the behavior of U3Si2 using available thermophysical data to predict the cladding-fuel pellet temperature and stress using the fuel performance code: BISON. The preliminary results obtained from the U3Si2 fuel model describe the mechanism of Pellet-Clad Mechanical Interaction for this system while more extensive testing including creep testing of U3Si2 is planned for improved understanding of thermophysical properties for predicting fuel performance.

  11. Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Worrall, Andrew [ORNL; Todosow, Michael [Brookhaven National Laboratory (BNL)

    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

  12. Basic data for integrated assessment of nuclear fuel cycle system

    Energy Technology Data Exchange (ETDEWEB)

    Nomura, Yasushi; Tamaki, Hitoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ito, Chihiro; Saegusa, Toshiari [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

    2001-03-01

    In our country, where natural energy resources such as oil and coal are scarce, it is vital to establish a nuclear fuel cycle to reprocess spent fuel and reuse valuable nuclear fuel in electric power generation reactors. However spent fuel is now being accumulated too much so that, for the time being, it is necessary to establish a system for tentatively storing spent fuel. In this report, in order to deal with these issues, evaluation methods, which were developed, prepared and discussed by Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI), are rendered together with sample results of their application. Also reported is some important information on the data and methods for the safety assessment of nuclear fuel cycle facilities, which have been surveyed by JAERI and CRIEPI. (author)

  13. Technoeconomy of different solid oxide fuel cell based hybrid cycle

    DEFF Research Database (Denmark)

    Rokni, Masoud

    2014-01-01

    Gas turbine, steam turbine and heat engine (Stirling engine) is used as bottoming cycle for a solid oxide fuel cell plant to compare different plants efficiencies, CO2 emissionsand plants cost in terms of $/kW. Each plant is then integrated with biomass gasification and finally six plants...... configurations are compared with each other. Technoeconomy is used when calculating the cost if the plants. It is found that when a solid oxide fuel cell plant is combined with a gas turbine cycle then the plant efficiency will be the highest one while if a biomass gasification plant is integrated...... with these hybrid cycles then integrated biomass gasification with solid oxide fuel cell and steam cycle will have the highest plant efficiency. The cost of solid oxide fuel cell with steam plant is found to be the lowest one with a value of about 1030$/kW....

  14. Fuel Cycle Services the Heart of Nuclear Energy

    Directory of Open Access Journals (Sweden)

    S. Soentono

    2007-01-01

    Full Text Available Fuel is essential for development whether for survival and or wealth creation purposes. In this century the utilization of fuels need to be improved although energy mix is still to be the most rational choice. The large amount utilization of un-renewable fossil has some disadvantages since its low energy content requires massive extraction, transport, and processing while emitting CO2 resulting degradation of the environment. In the mean time the advancement of nuclear science and technology has improved significantly the performance of nuclear power plant, management of radioactive waste, enhancement of proliferation resistance, and more economic competitiveness. Ever since the last decade of the last century the nuclear renaissance has taken place. This is also due to the fact that nuclear energy does not emit GHG. Although the nuclear fuel offers a virtually limitless source of economic energy, it is only so if the nuclear fuel is reprocessed and recycled. Consequently, the fuel cycle is to be even more of paramount important in the future. The infrastructure of the fuel cycle services worldwide has been adequately available. Various International Initiatives to access the fuel cycle services are also offered. However, it is required to put in place the International Arrangements to guaranty secured sustainable supply of services and its peaceful use. Relevant international co-operations are central for proceeding with the utilization of nuclear energy, while this advantageous nuclear energy utilization relies on the fuel cycle services. It is therefore concluded that the fuel cycle services are the heart of nuclear energy, and the international nuclear community should work together to maintain the availability of this nuclear fuel cycle services timely, sufficiently, and economically.

  15. Nuclear power generation and fuel cycle report 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-10-01

    This report presents the current status and projections through 2015 of nuclear capacity, generation, and fuel cycle requirements for all countries using nuclear power to generate electricity for commercial use. It also contains information and forecasts of developments in the worldwide nuclear fuel market. Long term projections of U.S. nuclear capacity, generation, and spent fuel discharges for two different scenarios through 2040 are developed. A discussion on decommissioning of nuclear power plants is included.

  16. Market-Based and System-Wide Fuel Cycle Optimization

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Paul Philip Hood [Univ. of Wisconsin, Madison, WI (United States); Scopatz, Anthony [Univ. of South Carolina, Columbia, SC (United States); Gidden, Matthew [Univ. of Wisconsin, Madison, WI (United States); Carlsen, Robert [Univ. of Wisconsin, Madison, WI (United States); Mouginot, Baptiste [Univ. of Wisconsin, Madison, WI (United States); Flanagan, Robert [Univ. of South Carolina, Columbia, SC (United States)

    2017-06-13

    This work introduces automated optimization into fuel cycle simulations in the Cyclus platform. This includes system-level optimizations, seeking a deployment plan that optimizes the performance over the entire transition, and market-level optimization, seeking an optimal set of material trades at each time step. These concepts were introduced in a way that preserves the flexibility of the Cyclus fuel cycle framework, one of its most important design principles.

  17. Assessment of the thorium fuel cycle in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Homan, F.J.; Allen, E.J.

    1977-01-01

    A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled.

  18. Wood energy fuel cycle optimization in beech and spruce forests

    Science.gov (United States)

    Meyer, Nickolas K.; Mina, Marco

    2012-03-01

    A novel synergistic approach to reducing emissions from residential wood combustion (RWC) is presented. Wood energy fuel cycle optimization (FCO) aims to provide cleaner burning fuels through optimization of forestry and renewable energy management practices. In this work, beech and spruce forests of average and high quality were modelled and analysed to determine the volume of fuel wood and its associated bark fraction produced during typical forestry cycles. Two separate fuel wood bark production regimes were observed for beech trees, while only one production regime was observed for spruce. The single tree and stand models were combined with existing thinning parameters to replicate existing management practices. Utilizing estimates of initial seedling numbers and existing thinning patterns a dynamic model was formed that responded to changes in thinning practices. By varying the thinning parameters, this model enabled optimization of the forestry practices for the reduction of bark impurities in the fuel wood supply chain. Beech forestry cycles responded well to fuel cycle optimization with volume reductions of bark from fuel wood of between ˜10% and ˜20% for average and high quality forest stands. Spruce, on the other hand, was fairly insensitive to FCO with bark reductions of 0-5%. The responsiveness of beech to FCO further supports its status as the preferred RWC fuel in Switzerland. FCO could easily be extended beyond Switzerland and applied across continental Europe and North America.

  19. Assessment of MARMOT. A Mesoscale Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Tonks, M. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Chakraborty, P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, X. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Fromm, B. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yu, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Teague, M. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Andersson, D. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    MARMOT is the mesoscale fuel performance code under development as part of the US DOE Nuclear Energy Advanced Modeling and Simulation Program. In this report, we provide a high level summary of MARMOT, its capabilities, and its current state of validation. The purpose of MARMOT is to predict the coevolution of microstructure and material properties of nuclear fuel and cladding. It accomplished this using the phase field method coupled to solid mechanics and heat conduction. MARMOT is based on the Multiphysics Object-Oriented Simulation Environment (MOOSE), and much of its basic capability in the areas of the phase field method, mechanics, and heat conduction come directly from MOOSE modules. However, additional capability specific to fuel and cladding is available in MARMOT. While some validation of MARMOT has been completed in the areas of fission gas behavior and grain growth, much more validation needs to be conducted. However, new mesoscale data needs to be obtained in order to complete this validation.

  20. RADIOACTIVE WASTE STREAMS FROM VARIOUS POTENTIAL NUCLEAR FUEL CYCLE OPTIONS

    Energy Technology Data Exchange (ETDEWEB)

    Nick Soelberg; Steve Piet

    2010-11-01

    Five fuel cycle options, about which little is known compared to more commonly known options, have been studied in the past year for the United States Department of Energy. These fuel cycle options, and their features relative to uranium-fueled light water reactor (LWR)-based fuel cycles, include: • Advanced once-through reactor concepts (Advanced Once-Through, or AOT) – intended for high uranium utilization and long reactor operating life, use depleted uranium in some cases, and avoid or minimize used fuel reprocessing • Fission-fusion hybrid (FFH) reactor concepts – potential variations are intended for high uranium or thorium utilization, produce fissile material for use in power generating reactors, or transmute transuranic (TRU) and some radioactive fission product (FP) isotopes • High temperature gas reactor (HTGR) concepts - intended for high uranium utilization, high reactor thermal efficiencies; they have unique fuel designs • Molten salt reactor (MSR) concepts – can breed fissile U-233 from Th fuel and avoid or minimize U fuel enrichment, use on-line reprocessing of the used fuel, produce lesser amounts of long-lived, highly radiotoxic TRU elements, and avoid fuel assembly fabrication • Thorium/U-233 fueled LWR (Th/U-233) concepts – can breed fissile U-233 from Th fuel and avoid or minimize U fuel enrichment, and produce lesser amounts of long-lived, highly radiotoxic TRU elements. These fuel cycle options could result in widely different types and amounts of used or spent fuels, spent reactor core materials, and waste streams from used fuel reprocessing, such as: • Highly radioactive, high-burnup used metal, oxide, or inert matrix U and/or Th fuels, clad in Zr, steel, or composite non-metal cladding or coatings • Spent radioactive-contaminated graphite, SiC, carbon-carbon-composite, metal, and Be reactor core materials • Li-Be-F salts containing U, TRU, Th, and fission products • Ranges of separated or un-separated activation

  1. Standalone BISON Fuel Performance Results for Watts Bar Unit 1, Cycles 1-3

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pawlowski, Roger [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Stimpson, Shane [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-03-07

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is moving forward with more complex multiphysics simulations and increased focus on incorporating fuel performance analysis methods. The coupled neutronics/thermal-hydraulics capabilities within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) have become relatively stable, and major advances have been made in analysis efforts, including the simulation of twelve cycles of Watts Bar Nuclear Unit 1 (WBN1) operation. While this is a major achievement, the VERA-CS approaches for treating fuel pin heat transfer have well-known limitations that could be eliminated through better integration with the BISON fuel performance code. Several approaches are being implemented to consider fuel performance, including a more direct multiway coupling with Tiamat, as well as a more loosely coupled one-way approach with standalone BISON cases. Fuel performance typically undergoes an independent analysis using a standalone fuel performance code with manually specified input defined from an independent core simulator solution or set of assumptions. This report summarizes the improvements made since the initial milestone to execute BISON from VERA-CS output. Many of these improvements were prompted through tighter collaboration with the BISON development team at Idaho National Laboratory (INL). A brief description of WBN1 and some of the VERA-CS data used to simulate it are presented. Data from a small mesh sensitivity study are shown, which helps justify the mesh parameters used in this work. The multi-cycle results are presented, followed by the results for the first three cycles of WBN1 operation, particularly the parameters of interest to pellet-clad interaction (PCI) screening (fuel-clad gap closure, maximum centerline fuel temperature, maximum/minimum clad hoop stress, and cumulative damage index). Once the mechanics of this capability are functioning, future work will target cycles with

  2. Criticality safety aspects of spent fuel arrays from emerging nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Nicolaou, G. [University of Thrace, Department of Electrical and Computer Engineering, Laboratory of Nuclear Technology, Kimmerria Campus, 67100 Xanthi (Greece)

    2010-07-01

    Emerging nuclear fuel cycles: fuels with Pu or minor actinides (MA) for their self-generated recycling or transmutation in PWR or FR {yields} reduction of radiotoxicity of HLW. The aim of work is to assess criticality (k{sub {infinity}}) of arrays of spent nuclear fuels from these emerging fuel cycles. Procedures: Calculations of - k{sub {infinity}}, using MCNP5 based on fresh and spent fuel compositions (infinite arrays), - spent fuel compositions using ORIGEN. Fuels considered: - commercial PWR-UO{sub 2} (R1) and -MOX (R2), [45 GWd/t] and fast reactor [100 GWd/t] (R3), - PWR self-generated Pu recycling (S1) and MA recycling (S2), FR self-generated MA recycling (S3), FR with 2% {sup 237}Np for transmutation purposes (T). Results: k{sub {infinity}} based on fresh and spent fuel compositions is shown. Fuels are clustered in two distinct families: - fast reactor fuels, - thermal reactor fuels; k{sub {infinity}} decreases when calculated on the basis of actinide and fission product inventory. In conclusions: - Emerging fuels considered resemble their corresponding commercial fuels; - k{sub {infinity}} decreases in all cases when calculated on the basis of spent fuel compositions (reactivity worth {approx}-20%{Delta}k/k), hence improving the effectiveness of packaging. (author)

  3. Economic Analysis of Different Nuclear Fuel Cycle Options

    Directory of Open Access Journals (Sweden)

    Won Il Ko

    2012-01-01

    Full Text Available An economic analysis has been performed to compare four nuclear fuel cycle options: a once-through cycle (OT, DUPIC recycling, thermal recycling using MOX fuel in a pressurized water reactor (PWR-MOX, and sodium fast reactor recycling employing pyroprocessing (Pyro-SFR. This comparison was made to suggest an economic competitive fuel cycle for the Republic of Korea. The fuel cycle cost (FCC has been calculated based on the equilibrium material flows integrated with the unit cost of the fuel cycle components. The levelized fuel cycle costs (LFCC have been derived in terms of mills/kWh for a fair comparison among the FCCs, and the results are as follows: OT 7.35 mills/kWh, DUPIC 9.06 mills/kWh, PUREX-MOX 8.94 mills/kWh, and Pyro-SFR 7.70 mills/kWh. Due to unavoidable uncertainties, a cost range has been applied to each unit cost, and an uncertainty study has been performed accordingly. A sensitivity analysis has also been carried out to obtain the break-even uranium price (215$/kgU for the Pyro-SFR against the OT, which demonstrates that the deployment of the Pyro-SFR may be economical in the foreseeable future. The influence of pyrotechniques on the LFCC has also been studied to determine at which level the potential advantages of Pyro-SFR can be realized.

  4. International nuclear fuel cycle fact book. Revision 4

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.

    1984-03-01

    This Fact Book has been compiled in an effort to provide (1) an overview of worldwide nuclear power and fuel cycle programs and (2) current data concerning fuel cycle and waste management facilities, R and D programs, and key personnel in countries other than the United States. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2. The Fact Book is organized as follows: (1) Overview section - summary tables which indicate national involvement in nuclear reactor, fuel cycle, and waste management development activities; (2) national summaries - a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; (3) international agencies - a section for each of the international agencies which has significant fuel cycle involvement; (4) energy supply and demand - summary tables, including nuclear power projections; (5) fuel cycle - summary tables; and (6) travel aids - international dialing instructions, international standard time chart, passport and visa requirements, and currency exchange rate.

  5. International Nuclear Fuel Cycle Fact Book. Revision 5

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.; Jeffs, A.G.

    1985-01-01

    This Fact Book has been compiled in an effort to provide: (1) an overview of worldwide nuclear power and fuel cycle programs; and (2) current data concerning fuel cycle and waste management facilities, R and D programs, and key personnel in countries other than the United States. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2. The Fact Book is organized as follows: (1) Overview section - summary tables which indicate national involvement in nuclear reactor, fuel cycle, and waste management development activities; (2) national summaries - a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; (3) international agencies - a section for each of the international agencies which has significant fuel cycle involvement; (4) energy supply and demand - summary tables, including nuclear power projections; (5) fuel cycle - summary tables; and (6) travel aids international dialing instructions, international standard time chart, passport and visa requirements, and currency exchange rate.

  6. Fuel Cycle Comparison for Distributed Power Technologies

    Energy Technology Data Exchange (ETDEWEB)

    Elgowainy, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Wang, M. Q. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2008-11-15

    This report examines backup power and prime power systems and addresses the potential energy and environmental effects of substituting fuel cells for existing combustion technologies based on microturbines and internal combustion engines.

  7. Advanced Nuclear Fuel Cycle Transitions: Optimization, Modeling Choices, and Disruptions

    Science.gov (United States)

    Carlsen, Robert W.

    Many nuclear fuel cycle simulators have evolved over time to help understan the nuclear industry/ecosystem at a macroscopic level. Cyclus is one of th first fuel cycle simulators to accommodate larger-scale analysis with it liberal open-source licensing and first-class Linux support. Cyclus also ha features that uniquely enable investigating the effects of modeling choices o fuel cycle simulators and scenarios. This work is divided into thre experiments focusing on optimization, effects of modeling choices, and fue cycle uncertainty. Effective optimization techniques are developed for automatically determinin desirable facility deployment schedules with Cyclus. A novel method fo mapping optimization variables to deployment schedules is developed. Thi allows relationships between reactor types and scenario constraints to b represented implicitly in the variable definitions enabling the usage o optimizers lacking constraint support. It also prevents wasting computationa resources evaluating infeasible deployment schedules. Deployed power capacit over time and deployment of non-reactor facilities are also included a optimization variables There are many fuel cycle simulators built with different combinations o modeling choices. Comparing results between them is often difficult. Cyclus flexibility allows comparing effects of many such modeling choices. Reacto refueling cycle synchronization and inter-facility competition among othe effects are compared in four cases each using combinations of fleet of individually modeled reactors with 1-month or 3-month time steps. There are noticeable differences in results for the different cases. The larges differences occur during periods of constrained reactor fuel availability This and similar work can help improve the quality of fuel cycle analysi generally There is significant uncertainty associated deploying new nuclear technologie such as time-frames for technology availability and the cost of buildin advanced reactors

  8. Shutdown Margin for High Conversion BWRs Operating in Th-233U Fuel Cycle

    CERN Document Server

    Shaposhnik, Yaniv; Elias, Ezra

    2013-01-01

    Several reactivity control system design options are explored in order to satisfy shutdown margin (SDM) requirements in a high conversion BWRs operating in Th-233U fuel cycle (Th-RBWR). The studied has an axially heterogeneous fuel assembly structure with a single fissile zone sandwiched between two fertile blanket zones. The utilization of an originally suggested RBWR Y-shape control rod in Th-RBWR is shown to be insufficient for maintaining adequate SDM to balance the high negative reactivity feedbacks, while maintaining fuel breeding potential, core power rating, and minimum Critical Power Ratio (CPR). Instead, an alternative assembly design, also relying on heterogeneous fuel zoning, is proposed for achieving fissile inventory ratio (FIR) above unity, adequate SDM and meeting minimum CPR limit at thermal core output matching the ABWR power. The new concept was modeled as a single 3-dimensional fuel assembly having reflective radial boundaries, using the BGCore system, which consists of the MCNP code coupl...

  9. Fuel cycle comparison of distributed power generation technologies.

    Energy Technology Data Exchange (ETDEWEB)

    Elgowainy, A.; Wang, M. Q.; Energy Systems

    2008-12-08

    The fuel-cycle energy use and greenhouse gas (GHG) emissions associated with the application of fuel cells to distributed power generation were evaluated and compared with the combustion technologies of microturbines and internal combustion engines, as well as the various technologies associated with grid-electricity generation in the United States and California. The results were primarily impacted by the net electrical efficiency of the power generation technologies and the type of employed fuels. The energy use and GHG emissions associated with the electric power generation represented the majority of the total energy use of the fuel cycle and emissions for all generation pathways. Fuel cell technologies exhibited lower GHG emissions than those associated with the U.S. grid electricity and other combustion technologies. The higher-efficiency fuel cells, such as the solid oxide fuel cell (SOFC) and molten carbonate fuel cell (MCFC), exhibited lower energy requirements than those for combustion generators. The dependence of all natural-gas-based technologies on petroleum oil was lower than that of internal combustion engines using petroleum fuels. Most fuel cell technologies approaching or exceeding the DOE target efficiency of 40% offered significant reduction in energy use and GHG emissions.

  10. World nuclear capacity and fuel cycle requirements, November 1993

    Energy Technology Data Exchange (ETDEWEB)

    1993-11-30

    This analysis report presents the current status and projections of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. Long-term projections of US nuclear capacity, generation, fuel cycle requirements, and spent fuel discharges for three different scenarios through 2030 are provided in support of the Department of Energy`s activities pertaining to the Nuclear Waste Policy Act of 1982 (as amended in 1987). The projections of uranium requirements also support the Energy Information Administration`s annual report, Domestic Uranium Mining and Milling Industry: Viability Assessment.

  11. Fusion fuel cycle: material requirements and potential effluents

    Energy Technology Data Exchange (ETDEWEB)

    Teofilo, V.L.; Bickford, W.E.; Long, L.W.; Price, B.A.; Mellinger, P.J.; Willingham, C.E.; Young, J.K.

    1980-10-01

    Environmental effluents that may be associated with the fusion fuel cycle are identified. Existing standards for controlling their release are summarized and anticipated regulatory changes are identified. The ability of existing and planned environmental control technology to limit effluent releases to acceptable levels is evaluated. Reference tokamak fusion system concepts are described and the principal materials required of the associated fuel cycle are analyzed. These materials include the fusion fuels deuterium and tritium; helium, which is used as a coolant for both the blanket and superconducting magnets; lithium and beryllium used in the blanket; and niobium used in the magnets. The chemical and physical processes used to prepare these materials are also described.

  12. Plant Characteristics of an Integrated Solid Oxide Fuel Cell Cycle and a Steam Cycle

    DEFF Research Database (Denmark)

    Rokni, Masoud

    2010-01-01

    Plant characteristics of a system containing a solid oxide fuel cell (SOFC) cycle on the top of a Rankine cycle were investigated. Natural gas (NG) was used as the fuel for the plant. A desulfurization reactor removes the sulfur content in the fuel, while a pre-reformer broke down the heavier...... hydrocarbons in an adiabatic steam reformer (ASR). The pre-treated fuel then entered to the anode side of the SOFC. The remaining fuels after the SOFC stacks entered a catalytic burner for further combusting. The burned gases from the burner were then used to produce steam for the Rankine cycle in a heat...... recovery steam generator (HRSG). The remaining energy of the off-gases was recycled back to the topping cycle for further utilization. Several parameter studies were carried out to investigate the sensitivity of the suggested plant. It was shown that the operation temperature of the desulfurization unit...

  13. Strategic research of advanced fuel cycle technologies in JNC

    Energy Technology Data Exchange (ETDEWEB)

    Kawata, T.; Fukushima, M.; Nomura, S. [Japan Nuclear Cycle Development Institute, Tokai Works (Japan)

    2000-07-01

    Key technologies for the future nuclear fuel cycle have been proposed and are being reviewed in JNC as a part of the Feasibility Study for an Advanced Fuel Cycle, which is to achieve a more flexible energy choice to satisfy a sustainable energy security and global environmental protection. The candidate reprocessing technologies are: 1) aqueous simplified PUREX process, 2) oxide or metallic electrowinning, and 3) fluoride volatilization for oxide, metal, or nitride fuels. The fuel fabrication methods being investigated are: 1) simplified pellet process, 2) sphere/vibro-packed process for MOX/MN fuel, and 3) casting for metal fuel. These candidate technologies are currently being compared based on past experiences, technical issues to be solved, industrial applicability for future plants, feasible options for MA/LLFP separation, and nonproliferation aspects. Alter two years of the present reviewing process, selected key technologies will be developed over the next five years to evaluate industrial applicability of reprocessing and fuel manufacturing processes for the advanced fuel cycle. (authors)

  14. Verify Super Double-Heterogeneous Spherical Lattice Model for Equilibrium Fuel Cycle Analysis AND HTR Spherical Super Lattice Model for Equilibrium Fuel Cycle Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gray S. Chang

    2005-11-01

    The currently being developed advanced High Temperature gas-cooled Reactors (HTR) is able to achieve a simplification of safety through reliance on innovative features and passive systems. One of the innovative features in these HTRs is reliance on ceramic-coated fuel particles to retain the fission products even under extreme accident conditions. Traditionally, the effect of the random fuel kernel distribution in the fuel pebble / block is addressed through the use of the Dancoff correction factor in the resonance treatment. However, the Dancoff correction factor is a function of burnup and fuel kernel packing factor, which requires that the Dancoff correction factor be updated during Equilibrium Fuel Cycle (EqFC) analysis. An advanced KbK-sph model and whole pebble super lattice model (PSLM), which can address and update the burnup dependent Dancoff effect during the EqFC analysis. The pebble homogeneous lattice model (HLM) is verified by the burnup characteristics with the double-heterogeneous KbK-sph lattice model results. This study summarizes and compares the KbK-sph lattice model and HLM burnup analyzed results. Finally, we discuss the Monte-Carlo coupling with a fuel depletion and buildup code - ORIGEN-2 as a fuel burnup analysis tool and its PSLM calculated results for the HTR EqFC burnup analysis.

  15. Technical ability of new MTR high-density fuel alloys regarding the whole fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Durand, J.P. [CERCA, Romans (France); Maugard, B. [CEA, F-75000 Paris (France); Gay, A. [Cogema, Velizy (France)

    1998-07-01

    The development of new fuel alloys could provide a good opportunity to improve drastically the fuel cycle on the neutronic performances and the reprocessing point of view. Nevertheless, those parameters can only be considered if the fuel manufacture feasibility has been previously demonstrated. As a matter of fact, a MTR work group involving French partners (CEA, CERCA, COGEMA) has been set up in order to evaluate the technical ability of new fuels considering the whole fuel cycle. In this paper CERCA is presenting the preliminary results of UMo and UNbZr fuel plate manufacture, CEA is comparing to U{sub 3}Si{sub 2} the neutronic performances of fuels such as UMo, UN, UNbZr, while COGEMA is dealing with the reprocessing feasibility. (author)

  16. Impact of the Taxes on Used Nuclear Fuel on the Fuel Cycle Economics in Spain

    Directory of Open Access Journals (Sweden)

    B. Yolanda Moratilla Soria

    2015-02-01

    Full Text Available In 2013, the Spanish government created two new taxes on used nuclear fuel. This article aims to present the results of an economic study carried out to compare the costs of long-term storage of used nuclear fuel –open cycle strategy–, with the cost of the strategy of reprocessing and recycling used fuel– closed cycle strategy– taking into account the impact of the new taxes on the global cost of the fuel cycle. The results show that the costs of open-cycle and closed-cycle spent fuel management, evaluated in Spain after the introduction of the taxes, are sufficiently similar (within the bounds of uncertainty, that the choice between both is predicated on other than purely economic criteria.

  17. The FIT Model - Fuel-cycle Integration and Tradeoffs

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Nick R. Soelberg; Samuel E. Bays; Candido Pereira; Layne F. Pincock; Eric L. Shaber; Meliisa C Teague; Gregory M Teske; Kurt G Vedros

    2010-09-01

    All mass streams from fuel separation and fabrication are products that must meet some set of product criteria – fuel feedstock impurity limits, waste acceptance criteria (WAC), material storage (if any), or recycle material purity requirements such as zirconium for cladding or lanthanides for industrial use. These must be considered in a systematic and comprehensive way. The FIT model and the “system losses study” team that developed it [Shropshire2009, Piet2010] are an initial step by the FCR&D program toward a global analysis that accounts for the requirements and capabilities of each component, as well as major material flows within an integrated fuel cycle. This will help the program identify near-term R&D needs and set longer-term goals. The question originally posed to the “system losses study” was the cost of separation, fuel fabrication, waste management, etc. versus the separation efficiency. In other words, are the costs associated with marginal reductions in separations losses (or improvements in product recovery) justified by the gains in the performance of other systems? We have learned that that is the wrong question. The right question is: how does one adjust the compositions and quantities of all mass streams, given uncertain product criteria, to balance competing objectives including cost? FIT is a method to analyze different fuel cycles using common bases to determine how chemical performance changes in one part of a fuel cycle (say used fuel cooling times or separation efficiencies) affect other parts of the fuel cycle. FIT estimates impurities in fuel and waste via a rough estimate of physics and mass balance for a set of technologies. If feasibility is an issue for a set, as it is for “minimum fuel treatment” approaches such as melt refining and AIROX, it can help to make an estimate of how performances would have to change to achieve feasibility.

  18. The FIT Model - Fuel-cycle Integration and Tradeoffs

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Nick R. Soelberg; Samuel E. Bays; Candido Pereira; Layne F. Pincock; Eric L. Shaber; Meliisa C Teague; Gregory M Teske; Kurt G Vedros

    2010-09-01

    All mass streams from fuel separation and fabrication are products that must meet some set of product criteria – fuel feedstock impurity limits, waste acceptance criteria (WAC), material storage (if any), or recycle material purity requirements such as zirconium for cladding or lanthanides for industrial use. These must be considered in a systematic and comprehensive way. The FIT model and the “system losses study” team that developed it [Shropshire2009, Piet2010] are an initial step by the FCR&D program toward a global analysis that accounts for the requirements and capabilities of each component, as well as major material flows within an integrated fuel cycle. This will help the program identify near-term R&D needs and set longer-term goals. The question originally posed to the “system losses study” was the cost of separation, fuel fabrication, waste management, etc. versus the separation efficiency. In other words, are the costs associated with marginal reductions in separations losses (or improvements in product recovery) justified by the gains in the performance of other systems? We have learned that that is the wrong question. The right question is: how does one adjust the compositions and quantities of all mass streams, given uncertain product criteria, to balance competing objectives including cost? FIT is a method to analyze different fuel cycles using common bases to determine how chemical performance changes in one part of a fuel cycle (say used fuel cooling times or separation efficiencies) affect other parts of the fuel cycle. FIT estimates impurities in fuel and waste via a rough estimate of physics and mass balance for a set of technologies. If feasibility is an issue for a set, as it is for “minimum fuel treatment” approaches such as melt refining and AIROX, it can help to make an estimate of how performances would have to change to achieve feasibility.

  19. Safety and Regulatory Issues of the Thorium Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian [ORNL; Worrall, Andrew [ORNL; Powers, Jeffrey [ORNL; Bowman, Steve [ORNL; Flanagan, George [ORNL; Gehin, Jess [ORNL

    2014-02-01

    Thorium has been widely considered an alternative to uranium fuel because of its relatively large natural abundance and its ability to breed fissile fuel (233U) from natural thorium (232Th). Possible scenarios for using thorium in the nuclear fuel cycle include use in different nuclear reactor types (light water, high temperature gas cooled, fast spectrum sodium, molten salt, etc.), advanced accelerator-driven systems, or even fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based on concepts that mix thorium with uranium (UO2 + ThO2), add fertile thorium (ThO2) fuel pins to LWR fuel assemblies, or use mixed plutonium and thorium (PuO2 + ThO2) fuel assemblies. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts on the nuclear fuel. Thorium and its irradiation products have nuclear characteristics that are different from those of uranium. In addition, ThO2, alone or mixed with UO2 fuel, leads to different chemical and physical properties of the fuel. These aspects are key to reactor safety-related issues. The primary objectives of this report are to summarize historical, current, and proposed uses of thorium in nuclear reactors; provide some important properties of thorium fuel; perform qualitative and quantitative evaluations of both in-reactor and out-of-reactor safety issues and requirements specific to a thorium-based fuel cycle for current LWR reactor designs; and identify key knowledge gaps and technical issues that need to be addressed for the licensing of thorium LWR fuel in the United States.

  20. Regulatory cross-cutting topics for fuel cycle facilities.

    Energy Technology Data Exchange (ETDEWEB)

    Denman, Matthew R.; Brown, Jason; Goldmann, Andrew Scott; Louie, David

    2013-10-01

    This report overviews crosscutting regulatory topics for nuclear fuel cycle facilities for use in the Fuel Cycle Research & Development Nuclear Fuel Cycle Evaluation and Screening study. In particular, the regulatory infrastructure and analysis capability is assessed for the following topical areas: Fire Regulations (i.e., how applicable are current Nuclear Regulatory Commission (NRC) and/or International Atomic Energy Agency (IAEA) fire regulations to advance fuel cycle facilities) Consequence Assessment (i.e., how applicable are current radionuclide transportation tools to support risk-informed regulations and Level 2 and/or 3 PRA) While not addressed in detail, the following regulatory topic is also discussed: Integrated Security, Safeguard and Safety Requirement (i.e., how applicable are current Nuclear Regulatory Commission (NRC) regulations to future fuel cycle facilities which will likely be required to balance the sometimes conflicting Material Accountability, Security, and Safety requirements.)

  1. Applying Advanced Neutron Transport Calculations for Improving Fuel Performance Codes

    Energy Technology Data Exchange (ETDEWEB)

    Botazzoli, P.; Luzzi, L. [Politecnico di Milano, Department of Energy, Nuclear Engineering Division - CeSNEF, Milano (Italy); Schubert, A.; Van Uffelen, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Karlsruhe (Germany); Haeck, W. [Institute de Radioprotection et de Surete Nucleaire, Fontenay-aux-Roses (France)

    2009-06-15

    TRANSURANUS is a computer code for the thermal and mechanical analysis of fuel rods in nuclear reactors. As part of the code, the TUBRNP model calculates the local concentration of the actinides (U, Pu, Am, Cm), the main fission products (Xe, Kr, Cs and Nd) and {sup 4}He produced during the irradiation as a function of the radial position across a fuel pellet (radial profiles). These local quantities are required for the determination of the local power density, the local burn-up, and the source term of fission products and other inert gases. In previous works the neutronic code ALEPH has been used to validate the models for the actinides and fission products concentrations in UO{sub 2} fuels. A similar approach has been adopted in the present work for verifying the Helium production. The present paper focuses on the modelling of the Helium production in PWR oxide fuels (MOX and UO{sub 2}). A reliable prediction of the Helium production and release in LWR oxide fuels is of great interest in case of increasing burn-up, linear heat generation rates and Plutonium content. The contribution of the Helium released plays a fundamental role in the gap pressure and subsequently in the mechanical behaviour of the fuel rod, in particular during the storage of the high burn-up spent fuel. Helium is produced in oxide fuels by three main paths: (i) alpha decay of the actinides (the main contribution is due to {sup 242}Cm, {sup 238}Pu and {sup 244}Cm); (ii) (n,{alpha}) reactions; and (iii) ternary fission. In the present work, the contributions due to ternary fission and the (n,{alpha}) reaction on {sup 16}O as well as some refinements in the {sup 241}Am burn-up chain have been included in TUBRNP. The VESTA neutronic code has been used for the validation of the He production model. The generic VESTA Monte Carlo depletion interface developed at IRSN allows us to couple different Monte Carlo codes with a depletion module. It currently allows for combining the ORIGEN 2.2 isotope

  2. Shutdown margin for high conversion BWRs operating in Th-{sup 233}U fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Shaposhnik, Y., E-mail: shaposhy@bgu.ac.il [NRCN – Nuclear Research Center Negev, POB 9001, Beer Sheva 84190 (Israel); Department of Nuclear Engineering, Ben-Gurion University of the Negev, POB 653, Beer Sheva 84105 (Israel); Shwageraus, E. [Department of Nuclear Engineering, Ben-Gurion University of the Negev, POB 653, Beer Sheva 84105 (Israel); Elias, E. [Faculty of Mechanical Engineering, Technion – Israel Institute of Technology, Technion City 32000, Haifa (Israel)

    2014-09-15

    Highlights: • BWR core operating in a closed self-sustainable Th-{sup 233}U fuel cycle. • Shutdown Margin in Th-RBWR design. • Fully coupled MC with fuel depletion and thermo-hydraulic feedback modules. • Thermal–hydraulic analysis includes MCPR observation. - Abstract: Several reactivity control system design options are explored in order to satisfy shutdown margin (SDM) requirements in a high conversion BWRs operating in Th-{sup 233}U fuel cycle (Th-RBWR). The studied core has an axially heterogeneous fuel assembly structure with a single fissile zone “sandwiched” between two fertile blanket zones. The utilization of an originally suggested RBWR Y-shape control rod in Th-RBWR is shown to be insufficient for maintaining adequate SDM to balance the high negative reactivity feedbacks, while maintaining fuel breeding potential, core power rating, and minimum Critical Power Ratio (CPR). Implementation of alternative reactivity control materials, reducing axial leakage through non-uniform enrichment distribution, use of burnable poisons, reducing number of pins as well as increasing pin diameter are also shown to be incapable of meeting the SDM requirements. Instead, an alternative assembly design, based on Rod Cluster Control Assembly with absorber rods was investigated. This design matches the reference ABWR core power and has adequate shutdown margin. The new concept was modeled as a single three-dimensional fuel assembly having reflective radial boundaries, using the BGCore system, which consists of the MCNP code coupled with fuel depletion and thermo-hydraulic feedback modules.

  3. Thermally regenerative hydrogen/oxygen fuel cell power cycles

    Science.gov (United States)

    Morehouse, J. H.

    1986-01-01

    Two innovative thermodynamic power cycles are analytically examined for future engineering feasibility. The power cycles use a hydrogen-oxygen fuel cell for electrical energy production and use the thermal dissociation of water for regeneration of the hydrogen and oxygen. The TDS (thermal dissociation system) uses a thermal energy input at over 2000 K to thermally dissociate the water. The other cycle, the HTE (high temperature electrolyzer) system, dissociates the water using an electrolyzer operating at high temperature (1300 K) which receives its electrical energy from the fuel cell. The primary advantages of these cycles is that they are basically a no moving parts system, thus having the potential for long life and high reliability, and they have the potential for high thermal efficiency. Both cycles are shown to be classical heat engines with ideal efficiency close to Carnot cycle efficiency. The feasibility of constructing actual cycles is investigated by examining process irreversibilities and device efficiencies for the two types of cycles. The results show that while the processes and devices of the 2000 K TDS exceed current technology limits, the high temperature electrolyzer system appears to be a state-of-the-art technology development. The requirements for very high electrolyzer and fuel cell efficiencies are seen as determining the feasbility of the HTE system, and these high efficiency devices are currently being developed. It is concluded that a proof-of-concept HTE system experiment can and should be conducted.

  4. A life-cycle comparison of alternative automobile fuels.

    Science.gov (United States)

    MacLean, H L; Lave, L B; Lankey, R; Joshi, S

    2000-10-01

    We examine the life cycles of gasoline, diesel, compressed natural gas (CNG), and ethanol (C2H5OH)-fueled internal combustion engine (ICE) automobiles. Port and direct injection and spark and compression ignition engines are examined. We investigate diesel fuel from both petroleum and biosources as well as C2H5OH from corn, herbaceous bio-mass, and woody biomass. The baseline vehicle is a gasoline-fueled 1998 Ford Taurus. We optimize the other fuel/powertrain combinations for each specific fuel as a part of making the vehicles comparable to the baseline in terms of range, emissions level, and vehicle lifetime. Life-cycle calculations are done using the economic input-output life-cycle analysis (EIO-LCA) software; fuel cycles and vehicle end-of-life stages are based on published model results. We find that recent advances in gasoline vehicles, the low petroleum price, and the extensive gasoline infrastructure make it difficult for any alternative fuel to become commercially viable. The most attractive alternative fuel is compressed natural gas because it is less expensive than gasoline, has lower regulated pollutant and toxics emissions, produces less greenhouse gas (GHG) emissions, and is available in North America in large quantities. However, the bulk and weight of gas storage cylinders required for the vehicle to attain a range comparable to that of gasoline vehicles necessitates a redesign of the engine and chassis. Additional natural gas transportation and distribution infrastructure is required for large-scale use of natural gas for transportation. Diesel engines are extremely attractive in terms of energy efficiency, but expert judgment is divided on whether these engines will be able to meet strict emissions standards, even with reformulated fuel. The attractiveness of direct injection engines depends on their being able to meet strict emissions standards without losing their greater efficiency. Biofuels offer lower GHG emissions, are sustainable, and

  5. Reducing Proliferation Rick Through Multinational Fuel Cycle Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Amanda Rynes

    2010-11-01

    With the prospect of rapid expansion of the nuclear energy industry and the ongoing concern over weapons proliferation, there is a growing need for a viable alternative to traditional nation-based fuel production facilities. While some in the international community remain apprehensive, the advantages of multinational fuel cycle facilities are becoming increasingly apparent, with states on both sides of the supply chain able to garner the security and financial benefits of such facilities. Proliferation risk is minimized by eliminating the need of states to establish indigenous fuel production capabilities and the concept's structure provides an additional internationally monitored barrier against the misuse or diversion of nuclear materials. This article gives a brief description of the arguments for and against the implementation of a complete multinational fuel cycle.

  6. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  7. Assessment of SFR fuel pin performance codes under advanced fuel for minor actinide transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Bouineau, V.; Lainet, M.; Chauvin, N.; Pelletier, M. [French Alternative Energies and Atomic Energy Commission - CEA, CEA Cadarache, DEN/DEC/SESC, 13108 Saint Paul lez Durance (France); Di Marcello, V.; Van Uffelen, P.; Walker, C. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Helmholtz-Platz 1, D- 76344 Eggenstein-Leopoldshafen (Germany)

    2013-07-01

    Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors of long-lived nuclides like {sup 241}Am is, therefore, an option for the reduction of radiotoxicity and residual power packages as well as the repository area. In the SUPERFACT Experiment four different oxide fuels containing high and low concentrations of {sup 237}Np and {sup 241}Am, representing the homogeneous and heterogeneous in-pile recycling concepts, were irradiated in the PHENIX reactor. The behavior of advanced fuel materials with minor actinide needs to be fully characterized, understood and modeled in order to optimize the design of this kind of fuel elements and to evaluate its performances. This paper assesses the current predictability of fuel performance codes TRANSURANUS and GERMINAL V2 on the basis of post irradiation examinations of the SUPERFACT experiment for pins with low minor actinide content. Their predictions have been compared to measured data in terms of geometrical changes of fuel and cladding, fission gases behavior and actinide and fission product distributions. The results are in good agreement with the experimental results, although improvements are also pointed out for further studies, especially if larger content of minor actinide will be taken into account in the codes. (authors)

  8. Life-cycle analysis of alternative aviation fuels in GREET

    Energy Technology Data Exchange (ETDEWEB)

    Elgowainy, A.; Han, J.; Wang, M.; Carter, N.; Stratton, R.; Hileman, J.; Malwitz, A.; Balasubramanian, S. (Energy Systems)

    2012-07-23

    The Greenhouse gases, Regulated Emissions, and Energy use in Transportation (GREET) model, developed at Argonne National Laboratory, has been expanded to include well-to-wake (WTWa) analysis of aviation fuels and aircraft. This report documents the key WTWa stages and assumptions for fuels that represent alternatives to petroleum jet fuel. The aviation module in GREET consists of three spreadsheets that present detailed characterizations of well-to-pump and pump-to-wake parameters and WTWa results. By using the expanded GREET version (GREET1{_}2011), we estimate WTWa results for energy use (total, fossil, and petroleum energy) and greenhouse gas (GHG) emissions (carbon dioxide, methane, and nitrous oxide) for (1) each unit of energy (lower heating value) consumed by the aircraft or (2) each unit of distance traveled/ payload carried by the aircraft. The fuel pathways considered in this analysis include petroleum-based jet fuel from conventional and unconventional sources (i.e., oil sands); Fisher-Tropsch (FT) jet fuel from natural gas, coal, and biomass; bio-jet fuel from fast pyrolysis of cellulosic biomass; and bio-jet fuel from vegetable and algal oils, which falls under the American Society for Testing and Materials category of hydroprocessed esters and fatty acids. For aircraft operation, we considered six passenger aircraft classes and four freight aircraft classes in this analysis. Our analysis revealed that, depending on the feedstock source, the fuel conversion technology, and the allocation or displacement credit methodology applied to co-products, alternative bio-jet fuel pathways have the potential to reduce life-cycle GHG emissions by 55-85 percent compared with conventional (petroleum-based) jet fuel. Although producing FT jet fuel from fossil feedstock sources - such as natural gas and coal - could greatly reduce dependence on crude oil, production from such sources (especially coal) produces greater WTWa GHG emissions compared with petroleum jet

  9. Life-Cycle Analysis of Alternative Aviation Fuels in GREET

    Energy Technology Data Exchange (ETDEWEB)

    Elgowainy, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Han, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Wang, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Carter, N. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Stratton, R. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hileman, J. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Malwitz, A. [Volpe National Transportation Systems Center, Cambridge, MA (United States); Balasubramanian, S. [Volpe National Transportation Systems Center, Cambridge, MA (United States)

    2012-06-01

    The Greenhouse gases, Regulated Emissions, and Energy use in Transportation (GREET) model, developed at Argonne National Laboratory, has been expanded to include well-to-wake (WTWa) analysis of aviation fuels and aircraft. This report documents the key WTWa stages and assumptions for fuels that represent alternatives to petroleum jet fuel. The aviation module in GREET consists of three spreadsheets that present detailed characterizations of well-to-pump and pump-to-wake parameters and WTWa results. By using the expanded GREET version (GREET1_2011), we estimate WTWa results for energy use (total, fossil, and petroleum energy) and greenhouse gas (GHG) emissions (carbon dioxide, methane, and nitrous oxide) for (1) each unit of energy (lower heating value) consumed by the aircraft or(2) each unit of distance traveled/ payload carried by the aircraft. The fuel pathways considered in this analysis include petroleum-based jet fuel from conventional and unconventional sources (i.e., oil sands); Fisher-Tropsch (FT) jet fuel from natural gas, coal, and biomass; bio-jet fuel from fast pyrolysis of cellulosic biomass; and bio-jet fuel from vegetable and algal oils, which falls under the American Society for Testing and Materials category of hydroprocessed esters and fatty acids. For aircraft operation, we considered six passenger aircraft classes and four freight aircraft classes in this analysis. Our analysis revealed that, depending on the feedstock source, the fuel conversion technology, and the allocation or displacement credit methodology applied to co-products, alternative bio-jet fuel pathways have the potential to reduce life-cycle GHG emissions by 55–85 percent compared with conventional (petroleum-based) jet fuel. Although producing FT jet fuel from fossil feedstock sources — such as natural gas and coal — could greatly reduce dependence on crude oil, production from such sources (especially coal) produces greater WTWa GHG emissions compared with petroleum jet

  10. Study on fission blanket fuel cycling of a fusion-fission hybrid energy generation system

    Science.gov (United States)

    Zhou, Z.; Yang, Y.; Xu, H.

    2011-10-01

    This paper presents a preliminary study on neutron physics characteristics of a light water cooled fission blanket for a new type subcritical fusion-fission hybrid reactor aiming at electric power generation with low technical limits of fission fuel. The major objective is to study the fission fuel cycling performance in the blanket, which may possess significant impacts on the feasibility of the new concept of fusion-fission hybrid reactor with a high energy gain (M) and tritium breeding ratio (TBR). The COUPLE2 code developed by the Institute of Nuclear and New Energy Technology of Tsinghua University is employed to simulate the neutronic behaviour in the blanket. COUPLE2 combines the particle transport code MCNPX with the fuel depletion code ORIGEN2. The code calculation results show that soft neutron spectrum can yield M > 20 while maintaining TBR >1.15 and the conversion ratio of fissile materials CR > 1 in a reasonably long refuelling cycle (>five years). The preliminary results also indicate that it is rather promising to design a high-performance light water cooled fission blanket of fusion-fission hybrid reactor for electric power generation by directly loading natural or depleted uranium if an ITER-scale tokamak fusion neutron source is achievable.

  11. Software Platform Evaluation - Verifiable Fuel Cycle Simulation (VISION) Model

    Energy Technology Data Exchange (ETDEWEB)

    J. J. Jacobson; D. E. Shropshire; W. B. West

    2005-11-01

    The purpose of this Software Platform Evaluation (SPE) is to document the top-level evaluation of potential software platforms on which to construct a simulation model that satisfies the requirements for a Verifiable Fuel Cycle Simulation Model (VISION) of the Advanced Fuel Cycle (AFC). See the Software Requirements Specification for Verifiable Fuel Cycle Simulation (VISION) Model (INEEL/EXT-05-02643, Rev. 0) for a discussion of the objective and scope of the VISION model. VISION is intended to serve as 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. This document will serve as a guide for selecting the most appropriate software platform for VISION. This is a “living document” that will be modified over the course of the execution of this work.

  12. International Nuclear Fuel Cycle Fact Book. Revision 12

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.

    1992-05-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need exists costs for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book has been compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NMEA activities reports; and proceedings of conferences and workshops. The data listed typically do not reflect any single source but frequently represent a consolidation/combination of information.

  13. International nuclear fuel cycle fact book. [Contains glossary

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.; Lakey, L.T.; Schneider, K.J.; Silviera, D.J.

    1987-01-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained has been obtained from nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops; and so forth. Sources do not agree completely with each other, and the data listed herein does not reflect any one single source but frequently is a consolidation/combination of information. Lack of space as well as the intent and purpose of the Fact Book limit the given information to that pertaining to the Nuclear Fuel Cycle and to data considered of primary interest or most helpful to the majority of users.

  14. Population exposure from the fuel cycle: Review and future direction

    Energy Technology Data Exchange (ETDEWEB)

    Richmond, C.R.

    1987-01-01

    The legacy of radiation exposures confronting man arises from two historical sources of energy, the sun and radioactive decay. Contemporary man continues to be dependent on these two energy sources, which include the nuclear fuel cycle. Radiation exposures from all energy sources should be examined, with particular emphasis on the nuclear fuel cycle, incidents such as Chernobyl and Three Mile Island. In addition to risk estimation, concepts such as de minimis, life shortening as a measure of risk, and competing risks as projected into the future must be considered in placing radiation exposures in perspective. The utility of these concepts is in characterizing population exposures for decision makers in a manner that the public may judge acceptable. All these viewpoints are essential in the evaluation of population exposure from the nuclear fuel cycle.

  15. International nuclear fuel cycle fact book: Revision 9

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.

    1989-01-01

    The International Nuclear Fuel Cycle Fact Book has been compiled in an effort to provide current data concerning fuel cycle and waste management facilities, R and D programs and key personnel. The Fact Book contains: national summaries in which a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; and international agencies in which a section for each of the international agencies which has significant fuel cycle involvement, and a listing of nuclear societies. The national summaries, in addition to the data described above, feature a small map for each country as well as some general information. The latter is presented from the perspective of the Fact Book user in the United States.

  16. Full fuel-cycle comparison of forklift propulsion systems.

    Energy Technology Data Exchange (ETDEWEB)

    Gaines, L. L.; Elgowainy, A.; Wang, M. Q.; Energy Systems

    2008-11-05

    Hydrogen has received considerable attention as an alternative to fossil fuels. The U.S. Department of Energy (DOE) investigates the technical and economic feasibility of promising new technologies, such as hydrogen fuel cells. A recent report for DOE identified three near-term markets for fuel cells: (1) Emergency power for state and local emergency response agencies, (2) Forklifts in warehousing and distribution centers, and (3) Airport ground support equipment markets. This report examines forklift propulsion systems and addresses the potential energy and environmental implications of substituting fuel-cell propulsion for existing technologies based on batteries and fossil fuels. Industry data and the Argonne Greenhouse Gases, Regulated Emissions, and Energy Use in Transportation (GREET) model are used to estimate full fuel-cycle emissions and use of primary energy sources, back to the primary feedstocks for fuel production. Also considered are other environmental concerns at work locations. The benefits derived from using fuel-cell propulsion are determined by the sources of electricity and hydrogen. In particular, fuel-cell forklifts using hydrogen made from the reforming of natural gas had lower impacts than those using hydrogen from electrolysis.

  17. Technology of the light water reactor fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Wymer, R. G.

    1979-01-01

    This essay presents elements of the processes used in the fuel cycle steps and gives an indication of the types of equipment used. The amounts of radioactivity released in normal operation of the processes are indicated and related to radiation doses. Types and costs of equipment or processes required to lower these radioactivity releases are in some cases suggested. Mining and milling, conversion of uranium concentrate to UF/sub 6/, uranium isotope separation, LWR fuel fabrication, fuel reprocessing, transportation, and waste management are covered in this essay. 40 figures, 34 tables. (DLC)

  18. Experimental validation of the DARWIN2.3 package for fuel cycle applications

    Energy Technology Data Exchange (ETDEWEB)

    San-Felice, L.; Eschbach, R.; Bourdot, P. [DEN, DER, CEA-Cadarache, F-13108 ST Paul-Lez-Durance (France); Tsilanizara, A.; Huynh, T. D. [DEN, DM2S, CEA-Saclay, F-91191 Gif sur Yvette (France); Ourly, H. [EDF, R and D, 1 av. General de Gaulle, F-92131 Clamart Cedex (France); Thro, J. F. [AREVA, Tour AREVA, F-92084 Paris la Defense (France)

    2012-07-01

    The DARWIN package, developed by the CEA and its French partners (AREVA and EDF) provides the required parameters for fuel cycle applications: fuel inventory, decay heat, activity, neutron, {gamma}, {alpha}, {beta} sources and spectrum, radiotoxicity. This paper presents the DARWIN2.3 experimental validation for fuel inventory and decay heat calculations on Pressurized Water Reactor (PWR). In order to validate this code system for spent fuel inventory a large program has been undertaken, based on spent fuel chemical assays. This paper deals with the experimental validation of DARWIN2.3 for the Pressurized Water Reactor (PWR) Uranium Oxide (UOX) and Mixed Oxide (MOX) fuel inventory calculation, focused on the isotopes involved in Burn-Up Credit (BUC) applications and decay heat computations. The calculation - experiment (C/E-1) discrepancies are calculated with the latest European evaluation file JEFF-3.1.1 associated with the SHEM energy mesh. An overview of the tendencies is obtained on a complete range of burn-up from 10 to 85 GWd/t (10 to 60 GWcVt for MOX fuel). The experimental validation of the DARWIN2.3 package for decay heat calculation is performed using calorimetric measurements carried out at the Swedish Interim Spent Fuel Storage Facility for Pressurized Water Reactor (PWR) assemblies, covering a large burn-up (20 to 50 GWd/t) and cooling time range (10 to 30 years). (authors)

  19. FRED fuel behaviour code: Main models and analysis of Halden IFA-503.2 tests

    Energy Technology Data Exchange (ETDEWEB)

    Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Shestopalov, A., E-mail: shest@dhtp.kiae.ru [RRC' Kurchatov Institute' , Kurchatov sq, 123182 Moscow (Russian Federation)

    2011-07-15

    Highlights: > We developed a new fuel rod behaviour code named FRED. > Main models and assumptions are described. > The code was checked using the IFA-503.2 tests performed at the Halden reactor. - Abstract: The FRED fuel rod code is being developed for thermal and mechanical simulation of fast breeder reactor (FBR) and light-water reactor (LWR) fuel behaviour under base-irradiation and accident conditions. The current version of the code calculates temperature distribution in fuel rods, stress-strain condition of cladding, fuel deformation, fuel-cladding gap conductance, and fuel rod inner pressure. The code was previously evaluated in the frame of two OECD mixed plutonium-uranium oxide (MOX) fuel performance benchmarks and then integrated into PSI's FAST code system to provide the fuel rod temperatures necessary for the neutron kinetics and thermal-hydraulic modules in transient calculations. This paper briefly overviews basic models and material property database of the FRED code used to assess the fuel behaviour under steady-state conditions. In addition, the code was used to simulate the IFA-503.2 tests, performed at the Halden reactor for two PWR and twelve VVER fuel samples under base-irradiation conditions. This paper presents the results of this simulation for two cases using a code-to-data comparison of fuel centreline temperatures, internal gas pressures, and fuel elongations. This comparison has demonstrated that the code adequately describes the important physical mechanisms of the uranium oxide (UOX) fuel rod thermal performance under steady-state conditions. Future activity should be concentrated on improving the model and extending the validation range, especially to the MOX fuel steady-state and transient behaviour.

  20. INTEGRATED GASIFICATION COMBINED CYCLE PROJECT 2 MW FUEL CELL DEMONSTRATION

    Energy Technology Data Exchange (ETDEWEB)

    FuelCell Energy

    2005-05-16

    With about 50% of power generation in the United States derived from coal and projections indicating that coal will continue to be the primary fuel for power generation in the next two decades, the Department of Energy (DOE) Clean Coal Technology Demonstration Program (CCTDP) has been conducted since 1985 to develop innovative, environmentally friendly processes for the world energy market place. The 2 MW Fuel Cell Demonstration was part of the Kentucky Pioneer Energy (KPE) Integrated Gasification Combined Cycle (IGCC) project selected by DOE under Round Five of the Clean Coal Technology Demonstration Program. The participant in the CCTDP V Project was Kentucky Pioneer Energy for the IGCC plant. FuelCell Energy, Inc. (FCE), under subcontract to KPE, was responsible for the design, construction and operation of the 2 MW fuel cell power plant. Duke Fluor Daniel provided engineering design and procurement support for the balance-of-plant skids. Colt Engineering Corporation provided engineering design, fabrication and procurement of the syngas processing skids. Jacobs Applied Technology provided the fabrication of the fuel cell module vessels. Wabash River Energy Ltd (WREL) provided the test site. The 2 MW fuel cell power plant utilizes FuelCell Energy's Direct Fuel Cell (DFC) technology, which is based on the internally reforming carbonate fuel cell. This plant is capable of operating on coal-derived syngas as well as natural gas. Prior testing (1992) of a subscale 20 kW carbonate fuel cell stack at the Louisiana Gasification Technology Inc. (LGTI) site using the Dow/Destec gasification plant indicated that operation on coal derived gas provided normal performance and stable operation. Duke Fluor Daniel and FuelCell Energy developed a commercial plant design for the 2 MW fuel cell. The plant was designed to be modular, factory assembled and truck shippable to the site. Five balance-of-plant skids incorporating fuel processing, anode gas oxidation, heat recovery

  1. Fuel-cycle assessment of selected bioethanol production.

    Energy Technology Data Exchange (ETDEWEB)

    Wu, M.; Wang, M.; Hong, H.; Energy Systems

    2007-01-31

    A large amount of corn stover is available in the U.S. corn belt for the potential production of cellulosic bioethanol when the production technology becomes commercially ready. In fact, because corn stover is already available, it could serve as a starting point for producing cellulosic ethanol as a transportation fuel to help reduce the nation's demand for petroleum oil. Using the data available on the collection and transportation of corn stover and on the production of cellulosic ethanol, we have added the corn stover-to-ethanol pathway in the GREET model, a fuel-cycle model developed at Argonne National Laboratory. We then analyzed the life-cycle energy use and emission impacts of corn stover-derived fuel ethanol for use as E85 in flexible fuel vehicles (FFVs). The analysis included fertilizer manufacturing, corn farming, farming machinery manufacturing, stover collection and transportation, ethanol production, ethanol transportation, and ethanol use in light-duty vehicles (LDVs). Energy consumption of petroleum oil and fossil energy, emissions of greenhouse gases (carbon dioxide [CO{sub 2}], nitrous oxide [N{sub 2}O], and methane [CH{sub 4}]), and emissions of criteria pollutants (carbon monoxide [CO], volatile organic compounds [VOCs], nitrogen oxide [NO{sub x}], sulfur oxide [SO{sub x}], and particulate matter with diameters smaller than 10 micrometers [PM{sub 10}]) during the fuel cycle were estimated. Scenarios of ethanol from corn grain, corn stover, and other cellulosic feedstocks were then compared with petroleum reformulated gasoline (RFG). Results showed that FFVs fueled with corn stover ethanol blends offer substantial energy savings (94-95%) relative to those fueled with RFG. For each Btu of corn stover ethanol produced and used, 0.09 Btu of fossil fuel is required. The cellulosic ethanol pathway avoids 86-89% of greenhouse gas emissions. Unlike the life cycle of corn grain-based ethanol, in which the ethanol plant consumes most of the fossil

  2. Economic Evaluation on the MOX Fuel in the Closed Fuel Cycle

    Directory of Open Access Journals (Sweden)

    Youqi Zheng

    2012-01-01

    Full Text Available The mixed oxide (MOX fuel is one of the most important fuels for the advanced reactors in the future. It is flexible to be applied either in the thermal reactor like pressurized water reactor (PWR or in the fast reactor (FR. This paper compares the two approaches from the view of fuel cost. Two features are involved. (1 The cost of electricity (COE is investigated based on the simulation of realistic operation of a practical PWR power plant and a typical fast breeder reactor design. (2 A new economic analysis model is established, considering the discount rate and the revenue of the reprocessed plutonium besides the traditional costs in the processes of fuel cycle. The sensitivity of COE to the changing parameters is also analyzed. The results show that, in the closed fuel cycle, the fuel cost of applying MOX fuels in the FBR is about 25% lower than that in the PWR at the current operating and fuel cycle level.

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

    Science.gov (United States)

    Palmer, David J.

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

  4. Symbiotic Pu/Th fuel cycle for HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Girardi, Enrico; Lecarpentier, David [EDF R and D, Les Renardieres, 77818 Moret sur Loing cedex (France)

    2006-07-01

    HTGRs, loaded with uranium or plutonium fuel do not authorize high breeding ratios. Consequently, the Gen IV requirement of sustainability with respect to the use of fissile resources may be difficult to meet. Relying on the thorium conversion into fissile uranium 233, the use of thorium is promising regarding the breeding ratio. Accordingly, the symbiotic Pu/Th fuel cycle is an interesting way to optimize the energetic use of plutonium. In the first part of this paper, we compare the performances of HTRs with different types of fuel (15% enriched U, Pu alone, Th-Pu) on the resource consumption. Relying on HTGRs, a good conversion ratio (typically 0.8) is mandatory to satisfy high temperature needs in the long term. In the second part, we discuss the neutronic calculation scheme for a prismatic HTGR annular core (GTMHR like, with an internal graphite reflector), and for a large HTGR (modeled with an assembly calculation). In the last part of the paper, Pu/Th, U233/Th and Pu-U233/Th fuel cycle are studied for the two types of HTGRs. The performances in terms of conversion ratios are compared. For a given power of 300 Mwe, several compositions for the fuel are studied. The fraction of thorium in the fuel composition is varying from 0% to 99% and the total fuel mass charged in the core ranges from 1200 kg to 15,000 kg. We show that good conversion ratios are very difficult to obtain for annular HTGRs. New designs must be considered, with a larger core, and a smaller power density (to reach the passive evacuation of residual power). Finally, a preliminary strategy for the deployment of these fuel cycles is presented. (authors)

  5. Open-Cycle Gas Turbine/Steam Turbine Combined Cycles with synthetic fuels from coal

    Science.gov (United States)

    Shah, R. P.; Corman, J. C.

    1977-01-01

    The Open-Cycle Gas Turbine/Steam Turbine Combined Cycle can be an effective energy conversion system for converting coal to electricity. The intermediate step in this energy conversion process is to convert the coal into a fuel acceptable to a gas turbine. This can be accomplished by producing a synthetic gas or liquid, and by removing, in the fuel conversion step, the elements in the fuel that would be harmful to the environment if combusted. In this paper, two open-cycle gas turbine combined systems are evaluated: one employing an integrated low-Btu gasifier, and one utilizing a semi-clean liquid fuel. A consistent technical/economic information base is developed for these two systems, and is compared with a reference steam plant burning coal directly in a conventional furnace.

  6. Simulation of Cycle-to-Cycle Variation in Dual-Fuel Engines

    KAUST Repository

    Jaasim, Mohammed

    2017-03-13

    Standard practices of internal combustion (IC) engine experiments are to conduct the measurements of quantities averaged over a large number of cycles. Depending on the operating conditions, the cycle-to-cycle variation (CCV) of quantities, such as the indicated mean effective pressure (IMEP) are observed at different levels. Accurate prediction of CCV in IC engines is an important but challenging task. Computational fluid dynamics (CFD) simulations using high performance computing (HPC) can be used effectively to visualize such 3D spatial distributions. In the present study, a dual fuel large engine is considered, with natural gas injected into the manifold accompanied with direct injection of diesel pilot fuel to trigger ignition. Multiple engine cycles in 3D are simulated in series as in the experiments to investigate the potential of HPC based high fidelity simulations to accurately capture the cycle to cycle variation in dual fuel engines. Open cycle simulations are conducted to predict the combined effect of the stratification of fuel-air mixture, temperature and turbulence on the CCV of pressure. The predicted coefficient of variation (COV) of pressure compared to the results from closed cycle simulations and the experiments.

  7. Thermodynamic Analysis of an Integrated Solid Oxide Fuel Cell Cycle with a Rankine Cycle

    DEFF Research Database (Denmark)

    Rokni, Masoud

    2010-01-01

    Hybrid systems consisting of Solid Oxide Fuel Cells (SOFC) on the top of a Steam Turbine (ST) are investigated. The plants are fired by natural gas (NG). A desulfurization reactor removes the sulfur content in the fuel while a pre-reformer breaks down the heavier hydrocarbons. The pre-treated fuel...... enters then into the anode side of the SOFC. The remaining fuels after the SOFC stacks enter a burner for further burning. The off-gases are then used to produce steam for a Rankine cycle in a Heat Recovery Steam Generator (HRSG). Different system setups are suggested. Cyclic efficiencies up to 67......% are achieved which is considerably higher than the conventional Combined Cycles (CC). Both ASR (Adiabatic Steam Reformer) and CPO (Catalytic Partial Oxidation) fuel pre-reformer reactors are considered in this investigation....

  8. Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

    Science.gov (United States)

    Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

    2016-09-01

    Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial (R-Z) or plane radial-circumferential (R-θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

  9. Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

    Science.gov (United States)

    Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

    2016-11-01

    Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial ( R- Z) or plane radial-circumferential ( R- θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

  10. VISION -- A Dynamic Model of the Nuclear Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    J. J. Jacobson; A. M. Yacout; S. J. Piet; D. E. Shropshire; G. E. Matthern

    2006-02-01

    The 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 deploy¬ment 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. VISION is intended to serve as 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.

  11. Estimating Externalities of Hydro Fuel Cycles, Report 6

    Energy Technology Data Exchange (ETDEWEB)

    Barnthouse, L.W.; Cada, G.F.; Cheng, M.-D.; Easterly, C.E.; Kroodsma, R.L.; Lee, R.; Shriner, D.S.; Tolbert, V.R.; Turner, R.S.

    1994-12-01

    There are three major objectives of this hydropower study: (1) to implement the methodological concepts that were developed in the background document (ORNL/RFF 1992) as a means of estimating the external costs and benefits of fuel cycles and, by so doing, to demonstrate their application to the hydroelectric fuel cycle (different fuel cycles have unique characteristics that need to be addressed in different ways); (2) to develop, given the time and resources, the best range of estimates of externalities associated with hydroelectric projects, using two benchmark projects at two reference sites in the US; and (3) to assess the state of the information that is available to support the estimation of externalities associated with the hydroelectric fuel cycle and, by so doing, to assist in identifying gaps in knowledge and in setting future research agendas. The main consideration in defining these objectives was a desire to have more information about externalities and a better method for estimating them. As set forth in the agreement between the US and the EC, the study is explicitly and intentionally not directed at any one audience. This study is about a methodology for estimating externalities. It is not about how to use estimates of externalities in a particular policy context.

  12. LMFBR operation in the nuclear cycle without fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Toshinsky, S.I. [Institute of Physics and Power Engineering, Kaluga (Russian Federation)

    1997-12-01

    Substantiation is given to expediency of investigation of nuclear power (NP) development with fast reactors cooled by lead-bismuth alloy operating during extended time in the open nuclear fuel cycle with slightly enriched or depleted uranium make-up. 9 refs., 1 fig., 6 tabs.

  13. Applicability of the SCALE code system to MOX fuel transport systems for criticality safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Toshihiro; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hayashi, Toshiaki; Takasugi, Masahiro; Natsume, Toshihiro; Tsuda, Kazuaki

    1996-11-01

    In order to ascertain feasibilities of the SCALE code system for MOX fuel transport systems, criticality analyses were performed for MOX fuel (Pu enrichment; 3.0 wt.%) criticality experiments at JAERI`s TCA and for infinite fuel rod arrays as parameters of Pu enrichment and lattice pitch. The comparison with a combination of the continuous energy Monte Carlo code MCNP and JENDL-3.2 indicated that the SCALE code system with GAM-THERMOS 123-group library can produce feasible results. Though HANSEN-ROACH 16-group library gives poorer results for MOS fuel transport systems, the errors are conservative except for high enriched fuels. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D. [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)

    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

  15. Moving toward multilateral mechanisms for the fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Panasyuk,A.; Rosenthal,M.; Efremov, G. V.

    2009-04-17

    Multilateral mechanisms for the fuel cycle are seen as a potentially important way to create an industrial infrastructure that will support a renaissance and at the same time not contribute to the risk of nuclear proliferation. In this way, international nuclear fuel cycle centers for enrichment can help to provide an assurance of supply of nuclear fuel that will reduce the likelihood that individual states will pursue this sensitive technology, which can be used to produce nuclear material directly usable nuclear weapons. Multinational participation in such mechanisms can also potentially promote transparency, build confidence, and make the implementation of IAEA safeguards more effective or more efficient. At the same time, it is important to ensure that there is no dissemination of sensitive technology. The Russian Federation has taken a lead role in this area by establishing an International Uranium Enrichment Center (IUEC) for the provision of enrichment services at its uranium enrichment plant located at the Angarsk Electrolysis Chemical Complex (AECC). This paper describes how the IUEe is organized, who its members are, and the steps that it has taken both to provide an assured supply of nuclear fuel and to ensure protection of sensitive technology. It also describes the relationship between the IUEC and the IAEA and steps that remain to be taken to enhance its assurance of supply. Using the IUEC as a starting point for discussion, the paper also explores more generally the ways in which features of such fuel cycle centers with multinational participation can have an impact on safeguards arrangements, transparency, and confidence-building. Issues include possible lAEA safeguards arrangements or other links to the IAEA that might be established at such fuel cycle centers, impact of location in a nuclear weapon state, and the transition by the IAEA to State Level safeguards approaches.

  16. Estimating Externalities of Natural Gas Fuel Cycles, Report 4

    Energy Technology Data Exchange (ETDEWEB)

    Barnthouse, L.W.; Cada, G.F.; Cheng, M.-D.; Easterly, C.E.; Kroodsma, R.L.; Lee, R.; Shriner, D.S.; Tolbert, V.R.; Turner, R.S.

    1998-01-01

    This report describes methods for estimating the external costs (and possibly benefits) to human health and the environment that result from natural gas fuel cycles. Although the concept of externalities is far from simple or precise, it generally refers to effects on individuals' well being, that result from a production or market activity in which the individuals do not participate, or are not fully compensated. In the past two years, the methodological approach that this report describes has quickly become a worldwide standard for estimating externalities of fuel cycles. The approach is generally applicable to any fuel cycle in which a resource, such as coal, hydro, or biomass, is used to generate electric power. This particular report focuses on the production activities, pollution, and impacts when natural gas is used to generate electric power. In the 1990s, natural gas technologies have become, in many countries, the least expensive to build and operate. The scope of this report is on how to estimate the value of externalities--where value is defined as individuals' willingness to pay for beneficial effects, or to avoid undesirable ones. This report is about the methodologies to estimate these externalities, not about how to internalize them through regulations or other public policies. Notwithstanding this limit in scope, consideration of externalities can not be done without considering regulatory, insurance, and other considerations because these institutional factors affect whether costs (and benefits) are in fact external, or whether they are already somehow internalized within the electric power market. Although this report considers such factors to some extent, much analysis yet remains to assess the extent to which estimated costs are indeed external. This report is one of a series of reports on estimating the externalities of fuel cycles. The other reports are on the coal, oil, biomass, hydro, and nuclear fuel cycles, and on general

  17. Cycle to Cycle Variation Study in a Dual Fuel Operated Engine

    KAUST Repository

    Pasunurthi, Shyamsundar

    2017-03-28

    The standard capability of engine experimental studies is that ensemble averaged quantities like in-cylinder pressure from multiple cycles and emissions are reported and the cycle to cycle variation (CCV) of indicated mean effective pressure (IMEP) is captured from many consecutive combustion cycles for each test condition. However, obtaining 3D spatial distribution of all the relevant quantities such as fuel-air mixing, temperature, turbulence levels and emissions from such experiments is a challenging task. Computational Fluid Dynamics (CFD) simulations of engine flow and combustion can be used effectively to visualize such 3D spatial distributions. A dual fuel engine is considered in the current study, with manifold injected natural gas (NG) and direct injected diesel pilot for ignition. Multiple engine cycles in 3D are simulated in series like in the experiments to investigate the potential of high fidelity RANS simulations coupled with detailed chemistry, to accurately predict the CCV. Cycle to cycle variation (CCV) is expected to be due to variabilities in operating and boundary conditions, in-cylinder stratification of diesel and natural gas fuels, variation in in-cylinder turbulence levels and velocity flow-fields. In a previous publication by the authors [1], variabilities in operating and boundary conditions are incorporated into several closed cycle simulations performed in parallel. Stochastic variations/stratifications of fuel-air mixture, turbulence levels, temperature and internal combustion residuals cannot be considered in such closed cycle simulations. In this study, open cycle simulations with port injection of natural gas predicted the combined effect of the stratifications on the CCV of in-cylinder pressure. The predicted Coefficient of Variation (COV) of cylinder pressure is improved compared to the one captured by closed cycle simulations in parallel.

  18. Reprocessed and combined thorium fuel cycles in a PER system with a micro heterogeneous approaches

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Fabiana B.A.; Castro, Victor F.; Faria, Rochkhudson B. de; Pereira, Claubia; Fortini, Angela, E-mail: fabianabeghini@yahoo.com.br, E-mail: victorfariacastro@gmail.com, E-mail: rochkdefaria@yahoo.com.br, E-mail: claubia@nuclear.ufmg.br, E-mail: fortini@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2015-07-01

    A micro heterogeneous approaches were used to study the behavior of reprocessed fuel spiked with thorium in a PWR fuel element considering (TRU-Th) cycle. The goal is to achieve a higher burnup using three different configurations to model the fuel element using SCALE 6.0. The reprocessed fuels were obtained using the ORIGEN 2.1 code from a spent PWR standard fuel (33,000 MWd/tHM burned), with 3.1% of initial enrichment. The spent fuel remained in the cooling pool for five years and then reprocessed using the UREX+ technique. Three configurations of micro heterogeneous approaches were analyzed, and the k{sub inf} and plutonium evolution during the burnup were evaluated. The preliminary results show that the behavior of advanced fuel based on transuranic elements spiked with thorium, and micro heterogeneous approach are satisfactory in PWRs, and the configuration that use a combination of Th and TRU (configuration 1) seems to be the most promising once has higher values for k{sub inf} during the burnup, compared with other configurations. (author)

  19. VISION User Guide - VISION (Verifiable Fuel Cycle Simulation) Model

    Energy Technology Data Exchange (ETDEWEB)

    Jacob J. Jacobson; Robert F. Jeffers; Gretchen E. Matthern; Steven J. Piet; Benjamin A. Baker; Joseph Grimm

    2009-08-01

    The purpose of this document is to provide a guide for using the current version of the Verifiable Fuel Cycle Simulation (VISION) model. This is a complex model with many parameters; the user is strongly encouraged to read this user guide before attempting to run the model. This model is an R&D work in progress and may contain errors and omissions. It is based upon numerous assumptions. This model is intended to assist in evaluating “what if” scenarios and in comparing fuel, reactor, and fuel processing alternatives at a systems level for U.S. nuclear power. The model is not intended as a tool for process flow and design modeling of specific facilities nor for tracking individual units of fuel or other material through the system. The model is intended to examine the interactions among the components of a fuel system as a function of time varying system parameters; this model represents a dynamic rather than steady-state approximation of the nuclear fuel system. VISION models the nuclear cycle at the system level, not individual facilities, e.g., “reactor types” not individual reactors and “separation types” not individual separation plants. Natural uranium can be enriched, which produces enriched uranium, which goes into fuel fabrication, and depleted uranium (DU), which goes into storage. Fuel is transformed (transmuted) in reactors and then goes into a storage buffer. Used fuel can be pulled from storage into either separation of disposal. If sent to separations, fuel is transformed (partitioned) into fuel products, recovered uranium, and various categories of waste. Recycled material is stored until used by its assigned reactor type. Note that recovered uranium is itself often partitioned: some RU flows with recycled transuranic elements, some flows with wastes, and the rest is designated RU. RU comes out of storage if needed to correct the U/TRU ratio in new recycled fuel. Neither RU nor DU are designated as wastes. VISION is comprised of several

  20. Thorium-Based Fuel Cycles in the Modular High Temperature Reactor

    Institute of Scientific and Technical Information of China (English)

    CHANG Hong; YANG Yongwei; JING Xingqing; XU Yunlin

    2006-01-01

    Large stockpiles of civil-grade as well as weapons-grade plutonium have been accumulated in the world from nuclear power or other programs of different countries. One alternative for the management of the plutonium is to incinerate it in the high temperature reactor (HTR). The thorium-based fuel cycle was studied in the modular HTR to reduce weapons-grade plutonium stockpiles, while producing no additional plutonium or other transuranic elements. Three thorium-uranium fuel cycles were also investigated. The thorium absorption cross sections of the resolved and unresolved resonances were generated using the ZUT-DGL code based on existing resonance data. The equilibrium core of the modular HTR was calculated and analyzed by means of the code VSOP'94. The results show that the modular HTR can incinerate most of the initially loaded plutonium amounting to about 95.3% net 239Pu for weapons-grade plutonium and can effectively utilize the uranium and thorium in the thorium-uranium fuel cycles.

  1. Evaluation of U-Zr hydride fuel for a thorium fuel cycle in an RTR concept

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Taek; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    In this paper, we performed a design study of a thorium fueled reactor according to the design concept of the Radkowsky Thorium Reactor (RTR) and evaluated its overall performance. To enhance its performance and alleviate its problems, we introduced a new metallic uranium fuel, uranium-zirconium hydride (U-ZrH{sub 1.6}), as a seed fuel. For comparison, typical ABB/CE-type PWR based on SYSTEM 80+and standard RTR-type thorium reactor were also studied. From the results of performance analysis, we could ascertain advantages of RTR-type thorium fueled reactor in proliferation resistance, fuel cycle economics, and back-end fuel cycle. Also, we found that enhancement of proliferation resistance and safer operating conditions may be achieved by using the U-ZrH{sub 1.6} fuel in the seed region without additional penalties in comparison with the standard RTR`s U-Zr fuel. 6 refs., 2 figs., 6 tabs. (Author)

  2. Thorium-based fuel cycles: Reassessment of fuel economics and proliferation risk

    Energy Technology Data Exchange (ETDEWEB)

    Serfontein, Dawid E., E-mail: Dawid.Serfontein@nwu.ac.za [Senior Lecturer at the School of Mechanical and Nuclear Engineering, North West University (PUK-Campus), PRIVATE BAG X6001, Internal Post Box 360, Potchefstroom 2520 (South Africa); Mulder, Eben J. [Professor at the School of Mechanical and Nuclear Engineering, North West University (South Africa)

    2014-05-01

    At current consumption and current prices, the proven reserves for natural uranium will last only about 100 years. However, the more abundant thorium, burned in breeder reactors, such as large High Temperature Gas-Cooled Reactors, and followed by chemical reprocessing of the spent fuel, could stretch the 100 years for uranium supply to 15,000 years. Thorium-based fuel cycles are also viewed as more proliferation resistant compared to uranium. However, several barriers to entry caused all countries, except India and Russia, to abandon their short term plans for thorium reactor projects, in favour of uranium/plutonium fuel cycles. In this article, based on the theory of resonance integrals and original analysis of fast fission cross sections, the breeding potential of {sup 232}Th is compared to that of {sup 238}U. From a review of the literature, the fuel economy of thorium-based fuel cycles is compared to that of natural uranium-based cycles. This is combined with a technical assessment of the proliferation resistance of thorium-based fuel cycles, based on a review of the literature. Natural uranium is currently so cheap that it contributes only about 10% of the cost of nuclear electricity. Chemical reprocessing is also very expensive. Therefore conservation of natural uranium by means of the introduction of thorium into the fuel is not yet cost effective and will only break even once the price of natural uranium were to increase from the current level of about $70/pound yellow cake to above about $200/pound. However, since fuel costs constitutes only a small fraction of the total cost of nuclear electricity, employing reprocessing in a thorium cycle, for the sake of its strategic benefits, may still be a financially viable option. The most important source of the proliferation resistance of {sup 232}Th/{sup 233}U fuel cycles is denaturisation of the {sup 233}U in the spent fuel by {sup 232}U, for which the highly radioactive decay chain potentially poses a large

  3. Fuel cycle analysis of once-through nuclear systems.

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. K.; Taiwo, T. A.; Nuclear Engineering Division

    2010-08-10

    Once-through fuel cycle systems are commercially used for the generation of nuclear power, with little exception. The bulk of these once-through systems have been water-cooled reactors (light-water and heavy water reactors, LWRs and HWRs). Some gas-cooled reactors are used in the United Kingdom. The commercial power systems that are exceptions use limited recycle (currently one recycle) of transuranic elements, primarily plutonium, as done in Europe and nearing deployment in Japan. For most of these once-through fuel cycles, the ultimate storage of the used (spent) nuclear fuel (UNF, SNF) will be in a geologic repository. Besides the commercial nuclear plants, new once-through concepts are being proposed for various objectives under international advanced nuclear fuel cycle studies and by industrial and venture capital groups. Some of the objectives for these systems include: (1) Long life core for remote use or foreign export and to support proliferation risk reduction goals - In these systems the intent is to achieve very long core-life with no refueling and limited or no access to the fuel. Most of these systems are fast spectrum systems and have been designed with the intent to improve plant economics, minimize nuclear waste, enhance system safety, and reduce proliferation risk. Some of these designs are being developed under Generation IV International Forum activities and have generally not used fuel blankets and have limited the fissile content of the fuel to less than 20% for the purpose on meeting international nonproliferation objectives. In general, the systems attempt to use transuranic elements (TRU) produced in current commercial nuclear power plants as this is seen as a way to minimize the amount of the problematic radio-nuclides that have to be stored in a repository. In this case, however, the reprocessing of the commercial LWR UNF to produce the initial fuel will be necessary. For this reason, some of the systems plan to use low enriched uranium

  4. Software Design Document for the AMP Nuclear Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Philip, Bobby [ORNL; Clarno, Kevin T [ORNL; Cochran, Bill [ORNL

    2010-03-01

    The purpose of this document is to describe the design of the AMP nuclear fuel performance code. It provides an overview of the decomposition into separable components, an overview of what those components will do, and the strategic basis for the design. The primary components of a computational physics code include a user interface, physics packages, material properties, mathematics solvers, and computational infrastructure. Some capability from established off-the-shelf (OTS) packages will be leveraged in the development of AMP, but the primary physics components will be entirely new. The material properties required by these physics operators include many highly non-linear properties, which will be replicated from FRAPCON and LIFE where applicable, as well as some computationally-intensive operations, such as gap conductance, which depends upon the plenum pressure. Because there is extensive capability in off-the-shelf leadership class computational solvers, AMP will leverage the Trilinos, PETSc, and SUNDIALS packages. The computational infrastructure includes a build system, mesh database, and other building blocks of a computational physics package. The user interface will be developed through a collaborative effort with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Capability Transfer program element as much as possible and will be discussed in detail in a future document.

  5. Globalisation of the nuclear fuel cycle - impact of developments on fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Durpel, L. van den; Bertel, E. [OECD Nuclear Energy Agency, 92 - Issy-les-Moulineaux (France)

    2000-02-01

    Nuclear energy will have to cope more and more with a rapid changing environment due to economic competitive pressure and the deregulatory progress. In current economic environment, utilities will have to focus strongly on the reduction of their total generation costs, covering the fuel cycle costs, which are only partly under their control. Developments in the fuel cycle will be in the short-term rather evolutionary addressing the current needs of utilities. However, within the context of sustainable development and more and more inclusion of externalities in energy generation costs, more performing developments in the fuel cycle could become important and feasible. A life-cycle design approach of the fuel cycle will be requested in order to cover all factors in order to decrease significantly the nuclear energy generation cost to complete with other alternative fuels in the long-term. This paper will report on some of the trends one could distinguish in the fuel cycle with emphasis on cost reduction. OECD/NEA is currently conducting a study on the fuel cycle aiming to assess current and future nuclear fuel cycles according to the potential for further improvement of the full added-value chain of these cycles from a mainly technological and economic perspective including environmental and social considerations. (orig.) [German] Die Kernenergie wird sich mehr und mehr in einem Umfeld behaupten muessen, das durch schnelle Veraenderungen auf Grund des Wettbewerbsdrucks in der Wirtschaft und des Liberalisierungsprozesses gekennzeichnet ist. Im heutigen Wirtschaftsumfeld muessen sich die Energieversorgungsunternehmen hauptsaechlich auf die Senkung ihrer Stromerzeugungs-Gesamtkosten konzentrieren. Darunter fallen auch die Brennstoffkreislaufkosten, die sie nur zum Teil beeinflussen koennen. Kurzfristig gesehen, duerften die Entwicklungen im Brennstoffkreislauf eher evolutionaer verlaufen und den jeweiligen Beduerfnissen der EVUs entsprechen. Im Zusammenhang mit einer

  6. Neutronic behavior of thorium fuel cycles in a very high temperature hybrid system

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

    Nuclear energy needs to guarantee four important issues to be successful as a sustainable energy source: nuclear safety, economic competitiveness, proliferation resistance and a minimal production of radioactive waste. Pebble bed reactors (PBR), which are very high temperature systems together with fuel cycles based in Thorium, they could offer the opportunity to meet the sustainability demands. Thorium is a potentially valuable energy source since it is about three to four times as abundant as Uranium. It is also a widely distributed natural resource readily accessible in many countries. This paper shows the main advantages of the use of a hybrid system formed by a Pebble Bed critical nuclear reactor and two Pebble Bed Accelerator Driven Systems (ADSs) using a variety of fuel cycles with Thorium (Th+U{sup 233}, Th+Pu{sup 239} and Th+U). The parameters related to the neutronic behavior like deep burn, nuclear fuel breeding, Minor Actinide stockpile, power density profiles and other are used to compare the fuel cycles using the well-known MCNPX computational code. (author)

  7. Validation of the BISON 3D Fuel Performance Code: Temperature Comparisons for Concentrically and Eccentrically Located Fuel Pellets

    Energy Technology Data Exchange (ETDEWEB)

    J. D. Hales; D. M. Perez; R. L. Williamson; S. R. Novascone; B. W. Spencer

    2013-03-01

    BISON is a modern finite-element based nuclear fuel performance code that has been under development at the Idaho National Laboratory (USA) since 2009. The code is applicable to both steady and transient fuel behaviour and is used to analyse either 2D axisymmetric or 3D geometries. BISON has been applied to a variety of fuel forms including LWR fuel rods, TRISO-coated fuel particles, and metallic fuel in both rod and plate geometries. Code validation is currently in progress, principally by comparison to instrumented LWR fuel rods. Halden IFA experiments constitute a large percentage of the current BISON validation base. The validation emphasis here is centreline temperatures at the beginning of fuel life, with comparisons made to seven rods from the IFA-431 and 432 assemblies. The principal focus is IFA-431 Rod 4, which included concentric and eccentrically located fuel pellets. This experiment provides an opportunity to explore 3D thermomechanical behaviour and assess the 3D simulation capabilities of BISON. Analysis results agree with experimental results showing lower fuel centreline temperatures for eccentric fuel with the peak temperature shifted from the centreline. The comparison confirms with modern 3D analysis tools that the measured temperature difference between concentric and eccentric pellets is not an artefact and provides a quantitative explanation for the difference.

  8. Life cycle models of conventional and alternative-fueled automobiles

    Science.gov (United States)

    Maclean, Heather Louise

    This thesis reports life cycle inventories of internal combustion engine automobiles with feasible near term fuel/engine combinations. These combinations include unleaded gasoline, California Phase 2 Reformulated Gasoline, alcohol and gasoline blends (85 percent methanol or ethanol combined with 15 percent gasoline), and compressed natural gas in spark ignition direct and indirect injection engines. Additionally, I consider neat methanol and neat ethanol in spark ignition direct injection engines and diesel fuel in compression ignition direct and indirect injection engines. I investigate the potential of the above options to have a lower environmental impact than conventional gasoline-fueled automobiles, while still retaining comparable pricing and consumer benefits. More broadly, the objective is to assess whether the use of any of the alternative systems will help to lead to the goal of a more sustainable personal transportation system. The principal tool is the Economic Input-Output Life Cycle Analysis model which includes inventories of economic data, environmental discharges, and resource use. I develop a life cycle assessment framework to assemble the array of data generated by the model into three aggregate assessment parameters; economics, externalities, and vehicle attributes. The first step is to develop a set of 'comparable cars' with the alternative fuel/engine combinations, based on characteristics of a conventional 1998 gasoline-fueled Ford Taurus sedan, the baseline vehicle for the analyses. I calculate the assessment parameters assuming that these comparable cars can attain the potential thermal efficiencies estimated by experts for each fuel/engine combination. To a first approximation, there are no significant differences in the assessment parameters for the vehicle manufacture, service, fixed costs, and the end-of-life for any of the options. However, there are differences in the vehicle operation life cycle components and the state of technology

  9. LIFE Materials: Fuel Cycle and Repository Volume 11

    Energy Technology Data Exchange (ETDEWEB)

    Shaw, H; Blink, J A

    2008-12-12

    The fusion-fission LIFE engine concept provides a path to a sustainable energy future based on safe, carbon-free nuclear power with minimal nuclear waste. The LIFE design ultimately offers many advantages over current and proposed nuclear energy technologies, and could well lead to a true worldwide nuclear energy renaissance. When compared with existing and other proposed future nuclear reactor designs, the LIFE engine exceeds alternatives in the most important measures of proliferation resistance and waste minimization. The engine needs no refueling during its lifetime. It requires no removal of fuel or fissile material generated in the LIFE engine. It leaves no weapons-attractive material at the end of life. Although there is certainly a need for additional work, all indications are that the 'back end' of the fuel cycle does not to raise any 'showstopper' issues for LIFE. Indeed, the LIFE concept has numerous benefits: (1) Per unit of electricity generated, LIFE engines would generate 20-30 times less waste (in terms of mass of heavy metal) requiring disposal in a HLW repository than does the current once-through fuel cycle. (2) Although there may be advanced fuel cycles that can compete with LIFE's low mass flow of heavy metal, all such systems require reprocessing, with attendant proliferation concerns; LIFE engines can do this without enrichment or reprocessing. Moreover, none of the advanced fuel cycles can match the low transuranic content of LIFE waste. (3) The specific thermal power of LIFE waste is initially higher than that of spent LWR fuel. Nevertheless, this higher thermal load can be managed using appropriate engineering features during an interim storage period, and could be accommodated in a Yucca-Mountain-like repository by appropriate 'staging' of the emplacement of waste packages during the operational period of the repository. The planned ventilation rates for Yucca Mountain would be sufficient for LIFE waste

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

    Science.gov (United States)

    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.

  11. Fuel utilization improvements in a once-through PWR fuel cycle. Final report on Task 6

    Energy Technology Data Exchange (ETDEWEB)

    Dabby, D.

    1979-06-01

    In studying the position of the United States Department of Energy, Non-proliferation Alternative Systems Assessment Program, this report determines the uranium saving associated with various improvement concepts applicable to a once-through fuel cycle of a standard four-loop Westinghouse Pressurized Water Reactor. Increased discharged fuel burnup from 33,000 to 45,000 MWD/MTM could achieve a 12% U/sub 3/O/sub 8/ saving by 1990. Improved fuel management schemes combined with coastdown to 60% power, could result in U/sub 3/O/sub 8/ savings of 6%.

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

    Science.gov (United States)

    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.

  13. The Adoption of Advanced Fuel Cycle Technology Under a Single Repository Policy

    Energy Technology Data Exchange (ETDEWEB)

    Paul Wilson

    2009-11-02

    Develops the tools to investiage the hypothesis that the savings in repository space associated with the implementation of advanced nuclear fuel cycles can result in sufficient cost savings to offset the higher costs of those fuel cycles.

  14. LIFE vs. LWR: End of the Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J C; Blink, J A; Shaw, H F

    2008-10-02

    LIFE are expected to result in a more straightforward licensing process and are also expected to improve the public perception of risk from nuclear power generation, transportation of nuclear materials, and nuclear waste disposal. Waste disposal is an ongoing issue for LWRs. The conventional (once-through) LWR fuel cycle treats unburned fuel as waste, and results in the current fleet of LWRs producing about twice as much waste in their 60 years of operation as is legally permitted to be disposed of in Yucca Mountain. Advanced LWR fuel cycles would recycle the unused fuel, such that each GWe-yr of electricity generation would produce only a small waste volume compared to the conventional fuel cycle. However, the advanced LWR fuel cycle requires chemical reprocessing plants for the fuel, multiple handling of radioactive materials, and an extensive transportation network for the fuel and waste. In contrast, the LIFE engine requires only one fueling for the plant lifetime, has no chemical reprocessing, and has a single shipment of a small amount of waste per GWe-yr of electricity generation. Public perception of the nuclear option will be improved by the reduction, for LIFE engines, of the number of shipments of radioactive material per GWe-yr and the need to build multiple repositories. In addition, LIFE fuel requires neither enrichment nor reprocessing, eliminating the two most significant pathways to proliferation from commercial nuclear fuel to weapons programs.

  15. The choice of the fuel assembly for VVER-1000 in a closed fuel cycle based on REMIX-technology

    Directory of Open Access Journals (Sweden)

    Bobrov Evgenii

    2016-01-01

    Full Text Available This paper shows basic features of different fuel assembly (FA application for VVER-1000 in a closed fuel cycle based on REMIX-technology. This investigation shows how the change in the water–fuel ratio in the VVER FA affects on the fuel characteristics produced by REMIX technology during multiple recycling.

  16. Reactor Fuel Isotopics and Code Validation for Nuclear Applications

    Energy Technology Data Exchange (ETDEWEB)

    Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Weber, Charles F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pigni, Marco T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-02-01

    Experimentally measured isotopic concentrations of well characterized spent nuclear fuel (SNF) samples have been collected and analyzed by previous researchers. These sets of experimental data have been used extensively to validate the accuracy of depletion code predictions for given sets of burnups, initial enrichments, and varying power histories for different reactor types. The purpose of this report is to present the diversity of data in a concise manner and summarize the current accuracy of depletion modeling. All calculations performed for this report were done using the Oak Ridge Isotope GENeration (ORIGEN) code, an internationally used irradiation and decay code solver within the SCALE comprehensive modeling and simulation code. The diversity of data given in this report includes key actinides, stable fission products, and radioactive fission products. In general, when using the current ENDF/B-VII.0 nuclear data libraries in SCALE, the major actinides are predicted to within 5% of the measured values. Large improvements were seen for several of the curium isotopes when using improved cross section data found in evaluated nuclear data file ENDF/B-VII.0 as compared to ENDF/B-V-based results. The impact of the flux spectrum on the plutonium isotope concentrations as a function of burnup was also shown. The general accuracy noted for the actinide samples for reactor types with burnups greater than 5,000 MWd/MTU was not observed for the low-burnup Hanford B samples. More work is needed in understanding these large discrepancies. The stable neodymium and samarium isotopes were predicted to within a few percent of the measured values. Large improvements were seen in prediction for a few of the samarium isotopes when using the ENDF/B-VII.0 libraries compared to results obtained with ENDF/B-V libraries. Very accurate predictions were obtained for 133Cs and 153Eu. However, the predicted values for the stable ruthenium and rhodium isotopes varied

  17. Development of DUPIC fuel cycle technology - Assessment of Wolsong NPP fuel handling system for DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Na, Bok Gyun; Nam, Gung Ihn [Korea Power Engineering Company, Taejon (Korea)

    2000-04-01

    The DUPIC fuel loading and discharge path of Wolsong NPP is studied assuming that DUPIC fuel is used at Wolsong NPP. Spent DUPIC fuel discharge path is irrelevant, since it uses the same spent fuel discharge path. Number of factors such as safety, economics of design change, radiation exposure to operators, easy of operation and maintenance, etc, are considered in the evaluation of path. A more detailed analysis of cost estimation of the selected path is also carried out. The study shows that DUPIC fuel loading path following through Spent Fuel Storage Bay and Spent Fuel Discharge Port in reverse direction will minimize the design change and additional equipment and radiation exposure to operators. The estimated total cost of using DUPIC fuel in Wolsong NPP based on price index of year 2000 is around 4.5 billion won. 4 refs., 30 figs., 13 tabs. (Author)

  18. Modification of Neutron Kinetic Code for Plate Type Fuel Nuclear Reactor

    Directory of Open Access Journals (Sweden)

    Salah Ud-Din Khan

    2013-01-01

    Full Text Available The research is conducted on the modification of neutron kinetic code for the plate type fuel nuclear reactor. REMARK is a neutron kinetic code that works only for the cylindrical type fuel nuclear reactor. In this research, our main emphasis is on the modification of this code in order to be applicable for the plate type fuel nuclear reactor. For this purpose, detailed mathematical studies have been performed and are subjected to write the program in Fortran language. Since REMARK code is written in Fortran language, so we have developed the program in Fortran and then inserted it into the source library of the code. The main emphasis is on the modification of subroutine in the source library of the code for hexagonal fuel assemblies with plate type fuel elements in it. The number of steps involved in the modification of the code has been included in the paper. The verification studies were performed by considering the small modular reactor with hexagonal assemblies and plate type fuel in it to find out the power distribution of the reactor core. The purpose of the research is to make the code work for the hexagonal fuel assemblies with plate type fuel element.

  19. Codes and Standards Requirements for Deployment of Emerging Fuel Cell Technologies

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, R.; Buttner, W.; Riykin, C.

    2011-12-01

    The objective of this NREL report is to provide information on codes and standards (of two emerging hydrogen power fuel cell technology markets; forklift trucks and backup power units), that would ease the implementation of emerging fuel cell technologies. This information should help project developers, project engineers, code officials and other interested parties in developing and reviewing permit applications for regulatory compliance.

  20. TEMP: a computer code to calculate fuel pin temperatures during a transient. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Bard, F E; Christensen, B Y; Gneiting, B C

    1980-04-01

    The computer code TEMP calculates fuel pin temperatures during a transient. It was developed to accommodate temperature calculations in any system of axi-symmetric concentric cylinders. When used to calculate fuel pin temperatures, the code will handle a fuel pin as simple as a solid cylinder or as complex as a central void surrounded by fuel that is broken into three regions by two circumferential cracks. Any fuel situation between these two extremes can be analyzed along with additional cladding, heat sink, coolant or capsule regions surrounding the fuel. The one-region version of the code accurately calculates the solution to two problems having closed-form solutions. The code uses an implicit method, an explicit method and a Crank-Nicolson (implicit-explicit) method.

  1. Estimating Externalities of Coal Fuel Cycles, Report 3

    Energy Technology Data Exchange (ETDEWEB)

    Barnthouse, L.W.; Cada, G.F.; Cheng, M.-D.; Easterly, C.E.; Kroodsma, R.L.; Lee, R.; Shriner, D.S.; Tolbert, V.R.; Turner, R.S.

    1994-09-01

    The agreement between the US DOE and the EC established the specific objectives of the study: (a) to develop a methodological framework that uses existing data and models to quantify the external costs and benefits of energy; (b) to demonstrate the application of the framework to estimate the externalities of the coal, biomass, oil, natural gas, hydro, nuclear, photovoltaic, and wind fuel cycles (by agreement with the EC, the US addressed the first six of these); and (c) to identify major gaps in the availability of information to quantify impacts, damages, benefits, and externalities of fuel cycles; and to suggest priorities for future research. The main consideration in defining these objectives was a desire to have more information about externalities, and a better method for estimating them.

  2. Selenium electrochemistry. Applications in the nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Maslennikov, A.; Peretroukhine, V. [Russian Academy of Sciences, Moscow (Russian Federation). Inst. of Physical Chemistry; David, F. [Centre National de la Recherche Scientifique (CNRS), 91 - Orsay (France); Lecomte, M. [CEA Centre d' Etudes de la Valle du Rhone, 30 - Marcoule (France). Direction du Cycle du Combustible

    1999-07-01

    Modern state of selenium electrochemistry is reviewed in respect of the application of electrochemical methods for the study of the behavior of this element and its quantitative analysis in the solutions of nuclear fuel cycle. The review includes the data on the redox potentials of Se in aqueous solutions, and the data on Se redox reactions, occurring at mercury and solid electrodes. Analysis of the available literature data shows that the inverse stripping voltammetry technique for trace Se concentration and determination seems to be the most promising in application for the Se determination in PUREX solutions and in radioactive wastes. The adaptation of the ISV technique for the trace Se concentration and determination in the solutions of the nuclear fuel cycle is indicated as the most prospective goal of the future experimental study. (author)

  3. Enduring Nuclear Fuel Cycle, Proceedings of a panel discussion

    Energy Technology Data Exchange (ETDEWEB)

    Walter, C. E., LLNL

    1997-11-18

    The panel reviewed the complete nuclear fuel cycle in the context of alternate energy resources, energy need projections, effects on the environment, susceptibility of nuclear materials to theft, diversion, and weapon proliferation. We also looked at ethical considerations of energy use, as well as waste, and its effects. The scope of the review extended to the end of the next century with due regard for world populations beyond that period. The intent was to take a long- range view and to project, not forecast, the future based on ethical rationales, and to avoid, as often happens, long-range discussions that quickly zoom in on only the next few decades. A specific nuclear fuel cycle technology that could satisfy these considerations was described and can be applied globally.

  4. A prospective fuel cycle for long lived radionuclide transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Jae Hyung; Shim, Joon Bo; Ahn, Byung Gil [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-07-01

    This study is aimed at outlining a prospective fuel cycle applicable to the nuclear transmutation of long lived radionuclides. Recycling the uranium and plutonium of spent fuels by Purex reprocessing is known to be far from an economical way at present. In addition, it generates high-level rad waste containing long-lived radionuclides, which would be a great burden to the subsequent final disposal in terms of cost, safety, and risks of environmental impact. As an alternative way to overcome this problem while still recycling the valuable fissionable materials as energy resources, transmutation is being taken into account by scientists in many countries. The concept of P{center_dot}T cycle suggested in this study is based on the technological requirements in relation with the transmutation system, while lowering the burden of waste disposal as well as risks of nuclear proliferation. (author)

  5. Analysis of advanced european nuclear fuel cycle scenarios including transmutation and economical estimates

    Energy Technology Data Exchange (ETDEWEB)

    Merino Rodriguez, I.; Alvarez-Velarde, F.; Martin-Fuertes, F. [CIEMAT, Avda. Complutense, 40, 28040 Madrid (Spain)

    2013-07-01

    In this work the transition from the existing Light Water Reactors (LWR) to the advanced reactors is analyzed, including Generation III+ reactors in a European framework. Four European fuel cycle scenarios involving transmutation options have been addressed. The first scenario (i.e., reference) is the current fleet using LWR technology and open fuel cycle. The second scenario assumes a full replacement of the initial fleet with Fast Reactors (FR) burning U-Pu MOX fuel. The third scenario is a modification of the second one introducing Minor Actinide (MA) transmutation in a fraction of the FR fleet. Finally, in the fourth scenario, the LWR fleet is replaced using FR with MOX fuel as well as Accelerator Driven Systems (ADS) for MA transmutation. All scenarios consider an intermediate period of GEN-III+ LWR deployment and they extend for a period of 200 years looking for equilibrium mass flows. The simulations were made using the TR-EVOL code, a tool for fuel cycle studies developed by CIEMAT. The results reveal that all scenarios are feasible according to nuclear resources demand (U and Pu). Concerning to no transmutation cases, the second scenario reduces considerably the Pu inventory in repositories compared to the reference scenario, although the MA inventory increases. The transmutation scenarios show that elimination of the LWR MA legacy requires on one hand a maximum of 33% fraction (i.e., a peak value of 26 FR units) of the FR fleet dedicated to transmutation (MA in MOX fuel, homogeneous transmutation). On the other hand a maximum number of ADS plants accounting for 5% of electricity generation are predicted in the fourth scenario (i.e., 35 ADS units). Regarding the economic analysis, the estimations show an increase of LCOE (Levelized cost of electricity) - averaged over the whole period - with respect to the reference scenario of 21% and 29% for FR and FR with transmutation scenarios respectively, and 34% for the fourth scenario. (authors)

  6. Characteriztion of particulate plutonium released in fuel cycle operations

    Energy Technology Data Exchange (ETDEWEB)

    Seefeldt, W.B.; Mecham, W.J.; Steindler, M.J.

    1976-05-01

    An estimate of the plutonium source terms is made for the fuel cycles of three reactor types on the basis of currently applied, currently available, and estimated future technology. The three reactors are LWR-U, LWR-Pu, and LMFBR. The source terms are characterized as to quantity, form, and particle size distribution. Historical operating data for existing plants and the state of the art of the technology of air cleaning are reviewed.

  7. Nuclear Fuel Cycle Reasoner: PNNL FY13 Report

    Energy Technology Data Exchange (ETDEWEB)

    Hohimer, Ryan E.; Strasburg, Jana D.

    2013-09-30

    In Fiscal Year 2012 (FY12) PNNL implemented a formal reasoning framework and applied it to a specific challenge in nuclear nonproliferation. The Semantic Nonproliferation Analysis Platform (SNAP) was developed as a preliminary graphical user interface to demonstrate the potential power of the underlying semantic technologies to analyze and explore facts and relationships relating to the nuclear fuel cycle (NFC). In Fiscal Year 2013 (FY13) the SNAP demonstration was enhanced with respect to query and navigation usability issues.

  8. System Losses Study - FIT (Fuel-cycle Integration and Tradeoffs)

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Nick R. Soelberg; Samuel E. Bays; Robert S. Cherry; Denia Djokic; Candido Pereira; Layne F. Pincock; Eric L. Shaber; Melissa C. Teague; Gregory M. Teske; Kurt G. Vedros

    2010-09-01

    This team aimed to understand the broad implications of changes of operating performance and parameters of a fuel cycle component on the entire system. In particular, this report documents the study of the impact of changing the loss of fission products into recycled fuel and the loss of actinides into waste. When the effort started in spring 2009, an over-simplified statement of the objective was “the number of nines” – how would the cost of separation, fuel fabrication, and waste management change as the number of nines of separation efficiency changed. The intent was to determine the optimum “losses” of TRU into waste for the single system that had been the focus of the Global Nuclear Energy Program (GNEP), namely sustained recycle in burner fast reactors, fed by transuranic (TRU) material recovered from used LWR UOX-51 fuel. That objective proved to be neither possible (insufficient details or attention to the former GNEP options, change in national waste management strategy from a Yucca Mountain focus) nor appropriate given the 2009-2010 change to a science-based program considering a wider range of options. Indeed, the definition of “losses” itself changed from the loss of TRU into waste to a generic definition that a “loss” is any material that ends up where it is undesired. All streams from either separation or fuel fabrication are products; fuel feed streams must lead to fuels with tolerable impurities and waste streams must meet waste acceptance criteria (WAC) for one or more disposal sites. And, these losses are linked in the sense that as the loss of TRU into waste is reduced, often the loss or carryover of waste into TRU or uranium is increased. The effort has provided a mechanism for connecting these three Campaigns at a technical level that had not previously occurred – asking smarter and smarter questions, sometimes answering them, discussing assumptions, identifying R&D needs, and gaining new insights. The FIT model has been a

  9. Evaluation Indicators for Analysis of Nuclear Fuel Cycle Sustainability

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Ko, Won Il; Chang, Hong Lae

    2008-01-15

    In this report, an attempt was made to derive indicators for the evaluation of the sustainability of the nuclear fuel cycle, using the methodologies developed by the INPRO, OECD/NEA and Gen-IV. In deriving the indicators, the three main elements of the sustainability, i.e., economics, environmental impact, and social aspect, as well as the technological aspect of the nuclear fuel cycle, considering the importance of the safety, were selected as the main criteria. An evaluation indicator for each criterion was determined, and the contents and evaluation method of each indicator were proposed. In addition, a questionnaire survey was carried out for the objectivity of the selection of the indicators in which participated some experts of the Korea Energy Technology and Emergency Management Institute (KETEMI) . Although the proposed indicators do not satisfy the characteristics and requirements of general indicators, it is presumed that they can be used in the analysis of the sustainability of the nuclear fuel cycle because those indicators incorporate various expert judgment and public opinions. On the other hand, the weighting factor of each indicator should be complemented in the future, using the AHP method and expert advice/consultations.

  10. A Concept of An Accelerator Closed Nuclear Fuel Cycle

    Science.gov (United States)

    Eremeev, I. P.

    1997-05-01

    The physical approach (I.P.Eremeev. Proc. of the PAC-95. Vol.1, p.98.) is applied for technology of nuclear fuel cycle. It is proposed the cycle to be closed by such an accelerator based process link, which would allow, on the one hand, the most hazardous of "equilibrium" radionuclides to be transmuted to stable isotopes or incinerated and, on the other hand, additional fissile fuel to be produced to compensate the energy consumption. Parameters of the technology, such as an intensity and energy "cost" of a transmutation event, a flux of photoneutrons produced have been determined for model targets. It is shown that the approach allows the above fission/transuranium radionuclides to be transmuted/ incinerated at a much greater rate than that of their build-up in operating NPP reactors at a much less energy consumption than an energy produced under their formation and at considerable compensation of the consumed energy through breeding fissile isotopes. A possibility of going to a closed Th-U fuel cycle is discussed. To realize the technology proposed requirements to a system of electron accelerators are formulated.

  11. A Deformation Analysis Code of CANDU Fuel under the Postulated Accident: ELOCA

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jung, Jong Yeob

    2006-11-15

    Deformations of the fuel element or fuel channel might be the main cause of the fuel failure. Therefore, the accurate prediction of the deformation and the analysis capabilities are closely related to the increase of the safety margin of the reactor. In this report, among the performance analysis or the transient behavior prediction computer codes, the analysis codes for deformation such as the ELOCA, HOTSPOT, CONTACT-1, and PTDFORM are briefly introduced and each code's objectives, applicability, and relations are explained. Especially, the user manual for ELOCA code which is the analysis code for the fuel deformation and the release of fission product during the transient period after the postulated accidents is provided so that it can be the guidance to the potential users of the code and save the time and economic loss by reducing the trial and err000.

  12. Estimating Fuel Cycle Externalities: Analytical Methods and Issues, Report 2

    Energy Technology Data Exchange (ETDEWEB)

    Barnthouse, L.W.; Cada, G.F.; Cheng, M.-D.; Easterly, C.E.; Kroodsma, R.L.; Lee, R.; Shriner, D.S.; Tolbert, V.R.; Turner, R.S.

    1994-07-01

    The activities that produce electric power typically range from extracting and transporting a fuel, to its conversion into electric power, and finally to the disposition of residual by-products. This chain of activities is called a fuel cycle. A fuel cycle has emissions and other effects that result in unintended consequences. When these consequences affect third parties (i.e., those other than the producers and consumers of the fuel-cycle activity) in a way that is not reflected in the price of electricity, they are termed ''hidden'' social costs or externalities. They are the economic value of environmental, health and any other impacts, that the price of electricity does not reflect. How do you estimate the externalities of fuel cycles? Our previous report describes a methodological framework for doing so--called the damage function approach. This approach consists of five steps: (1) characterize the most important fuel cycle activities and their discharges, where importance is based on the expected magnitude of their externalities, (2) estimate the changes in pollutant concentrations or other effects of those activities, by modeling the dispersion and transformation of each pollutant, (3) calculate the impacts on ecosystems, human health, and any other resources of value (such as man-made structures), (4) translate the estimates of impacts into economic terms to estimate damages and benefits, and (5) assess the extent to which these damages and benefits are externalities, not reflected in the price of electricity. Each step requires a different set of equations, models and analysis. Analysts generally believe this to be the best approach for estimating externalities, but it has hardly been used! The reason is that it requires considerable analysis and calculation, and to this point in time, the necessary equations and models have not been assembled. Equally important, the process of identifying and estimating externalities leads to a number

  13. Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

    Science.gov (United States)

    Alekseev, P. N.; Bobrov, E. A.; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A.

    2015-12-01

    The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U-Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium-plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: 235U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or 233U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

  14. Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, P. N.; Bobrov, E. A., E-mail: evgeniybobrov89@rambler.ru; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A. [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U–Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium–plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: {sup 235}U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or {sup 233}U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

  15. Life cycle assessment of automobile/fuel options.

    Science.gov (United States)

    MacLean, Heather L; Lave, Lester B

    2003-12-01

    We examine the possibilities for a "greener" car that would use less material and fuel, be less polluting, and would have a well-managed end-of-life. Light-duty vehicles are fundamental to our economy and will continue to be for the indefinite future. Any redesign to make these vehicles greener requires consumer acceptance. Consumer desires for large, powerful vehicles have been the major stumbling block in achieving a "green car". The other major barrier is inherent contradictions among social goals such as fuel economy, safety, low emissions of pollutants, and low emissions of greenhouse gases, which has led to conflicting regulations such as emissions regulations blocking sales of direct injection diesels in California, which would save fuel. In evaluating fuel/vehicle options with the potential to improve the greenness of cars [diesel (direct injection) and ethanol in internal combustion engines, battery-powered, gasoline hybrid electric, and hydrogen fuel cells], we find no option dominates the others on all dimensions. The principles of green design developed by Anastas and Zimmerman (Environ. Sci. Technol. 2003, 37, 94A-101A) and the use of a life cycle approach provide insights on the key sustainability issues associated with the various options.

  16. Projections of Full-Fuel-Cycle Energy and Emissions Metrics

    Energy Technology Data Exchange (ETDEWEB)

    Coughlin, Katie [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    2013-01-01

    To accurately represent how conservation and efficiency policies affect energy demand, both direct and indirect impacts need to be included in the accounting. The indirect impacts are defined here as the resource savings that accrue over the fuel production chain, which when added to the energy consumed at the point of use, constitute the full-fuel- cycle (FFC) energy. This paper uses the accounting framework developed in (Coughlin 2012) to calculate FFC energy metrics as time series for the period 2010-2040. The approach is extended to define FFC metrics for the emissions of greenhouse gases (GHGs) and other air-borne pollutants. The primary focus is the types of energy used in buildings and industrial processes, mainly natural gas and electricity. The analysis includes a discussion of the fuel production chain for coal, which is used extensively for electric power generation, and for diesel and fuel oil, which are used in mining, oil and gas operations, and fuel distribution. Estimates of the energy intensity parameters make use of data and projections from the Energy Information Agency’s National Energy Modeling System, with calculations based on information from the Annual Energy Outlook 2012.

  17. 40 CFR 600.114-08 - Vehicle-specific 5-cycle fuel economy calculations.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Vehicle-specific 5-cycle fuel economy... (CONTINUED) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy... fuel economy calculations. This section applies to data used for fuel economy labeling under Subpart D...

  18. Computer code for single-point thermodynamic analysis of hydrogen/oxygen expander-cycle rocket engines

    Science.gov (United States)

    Glassman, Arthur J.; Jones, Scott M.

    1991-01-01

    This analysis and this computer code apply to full, split, and dual expander cycles. Heat regeneration from the turbine exhaust to the pump exhaust is allowed. The combustion process is modeled as one of chemical equilibrium in an infinite-area or a finite-area combustor. Gas composition in the nozzle may be either equilibrium or frozen during expansion. This report, which serves as a users guide for the computer code, describes the system, the analysis methodology, and the program input and output. Sample calculations are included to show effects of key variables such as nozzle area ratio and oxidizer-to-fuel mass ratio.

  19. Fuel cycle analysis of once-through nuclear systems.

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. K.; Taiwo, T. A.; Nuclear Engineering Division

    2010-08-10

    Once-through fuel cycle systems are commercially used for the generation of nuclear power, with little exception. The bulk of these once-through systems have been water-cooled reactors (light-water and heavy water reactors, LWRs and HWRs). Some gas-cooled reactors are used in the United Kingdom. The commercial power systems that are exceptions use limited recycle (currently one recycle) of transuranic elements, primarily plutonium, as done in Europe and nearing deployment in Japan. For most of these once-through fuel cycles, the ultimate storage of the used (spent) nuclear fuel (UNF, SNF) will be in a geologic repository. Besides the commercial nuclear plants, new once-through concepts are being proposed for various objectives under international advanced nuclear fuel cycle studies and by industrial and venture capital groups. Some of the objectives for these systems include: (1) Long life core for remote use or foreign export and to support proliferation risk reduction goals - In these systems the intent is to achieve very long core-life with no refueling and limited or no access to the fuel. Most of these systems are fast spectrum systems and have been designed with the intent to improve plant economics, minimize nuclear waste, enhance system safety, and reduce proliferation risk. Some of these designs are being developed under Generation IV International Forum activities and have generally not used fuel blankets and have limited the fissile content of the fuel to less than 20% for the purpose on meeting international nonproliferation objectives. In general, the systems attempt to use transuranic elements (TRU) produced in current commercial nuclear power plants as this is seen as a way to minimize the amount of the problematic radio-nuclides that have to be stored in a repository. In this case, however, the reprocessing of the commercial LWR UNF to produce the initial fuel will be necessary. For this reason, some of the systems plan to use low enriched uranium

  20. Cycle update : advanced fuels and technologies for emissions reduction

    Energy Technology Data Exchange (ETDEWEB)

    Smallwood, G. [National Research Council of Canada, Ottawa, ON (Canada)

    2009-07-01

    This paper provided a summary of key achievements of the Program of Energy Research and Development advanced fuels and technologies for emissions reduction (AFTER) program over the funding cycle from fiscal year 2005/2006 to 2008/2009. The purpose of the paper was to inform interested parties of recent advances in knowledge and in science and technology capacities in a concise manner. The paper discussed the high level research and development themes of the AFTER program through the following 4 overarching questions: how could advanced fuels and internal combustion engine designs influence emissions; how could emissions be reduced through the use of engine hardware including aftertreatment devices; how do real-world duty cycles and advanced technology vehicles operating on Canadian fuels compare with existing technologies, models and estimates; and what are the health risks associated with transportation-related emissions. It was concluded that the main issues regarding the use of biodiesel blends in current technology diesel engines are the lack of consistency in product quality; shorter shelf life of biodiesel due to poorer oxidative stability; and a need to develop characterization methods for the final oxygenated product because most standard methods are developed for hydrocarbons and are therefore inadequate. 2 tabs., 13 figs.

  1. Safety research in nuclear fuel cycle at PNC

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    This report collects the results of safety research in nuclear fuel cycle at Power Reactor and Nuclear Fuel Development Corporation, in order to answer to the Questionnaire of OECD/NEA. The Questionnaire request to include information concerning to research topic, description, main results (if available), reference documents, research institutes involved, sponsoring organization and other pertinent information about followings: a) Recently completed research projects. b) Ongoing (current) research projects. Achievements on following items are omitted by the request of OECD/NEA, uranium mining and milling, uranium refining and conversion to UF{sub 6}, uranium enrichment, fuel manufacturers, spent fuel storage, radioactive waste management, transport of radioactive materials, decommissioning. We select topics from the fields of a) nuclear installation, b) seismic, and c) PSA, in projects from frame of annual safety research plan for nuclear installations established by Nuclear Safety Commission. We apply for the above a) and b) projects as follows: a) Achievements in Safety Research, fiscal 1991-1995, b) fiscal 1996 Safety Research Achievements: Progress. (author)

  2. Safety research in nuclear fuel cycle at PNC

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    This report collects the results of safety research in nuclear fuel cycle at Power Reactor and Nuclear Fuel Development Corporation, in order to answer to the Questionnaire of OECD/NEA. The Questionnaire request to include information concerning to research topic, description, main results (if available), reference documents, research institutes involved, sponsoring organization and other pertinent information about followings: a) Recently completed research projects. b) Ongoing (current) research projects. Achievements on following items are omitted by the request of OECD/NEA, uranium mining and milling, uranium refining and conversion to UF{sub 6}, uranium enrichment, fuel manufacturers, spent fuel storage, radioactive waste management, transport of radioactive materials, decommissioning. We select topics from the fields of a) nuclear installation, b) seismic, and c) PSA, in projects from frame of annual safety research plan for nuclear installations established by Nuclear Safety Commission. We apply for the above a) and b) projects as follows: a) Achievements in Safety Research, fiscal 1991-1995, b) fiscal 1996 Safety Research Achievements: Progress. (author)

  3. Novel BCH Code Design for Mitigation of Phase Noise Induced Cycle Slips in DQPSK Systems

    DEFF Research Database (Denmark)

    Leong, M. Y.; Larsen, Knud J.; Jacobsen, G.

    2014-01-01

    We show that by proper code design, phase noise induced cycle slips causing an error floor can be mitigated for 28 Gbau d DQPSK systems. Performance of BCH codes are investigated in terms of required overhead......We show that by proper code design, phase noise induced cycle slips causing an error floor can be mitigated for 28 Gbau d DQPSK systems. Performance of BCH codes are investigated in terms of required overhead...

  4. Developing the User Experience for a Next Generation Nuclear Fuel Cycle Simulator (NGFCS)

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Paul H. [Univ. of Wisconsin, Madison, WI (United States); Schneider, Erich [Univ. of Texas, Austin, TX (United States); Pascucci, Valerio [Univ. of Utah, Salt Lake City, UT (United States); Livnat, Yarden [Univ. of Utah, Salt Lake City, UT (United States); Hiromoto, Robert [Univ. of Idaho, Moscow, ID (United States); Scopatz, Anthony [Univ. of Wisconsin, Madison, WI (United States); Brossard, Dominique [Univ. of Wisconsin, Madison, WI (United States); Scheufele, Dietram [Univ. of Wisconsin, Madison, WI (United States)

    2017-07-30

    This project made substantial progress on its original aim for providing a modern user experience for nuclear fuel cycle analysis while also creating a robust and functional next- generation fuel cycle simulator. The Cyclus kernel experienced a dramatic clari cation of its interfaces and data model, becoming a full- edged agent-based framework, with strong support for third party developers of novel archetypes. The most important contribution of this project to the the development of Cyclus was the introduction of tools to facilitate archetype development. These include automated code generation of routine archetype components, metadata annotations to provide re ection and rich description of each data member's purpose, and mechanisms for input validation and output of complex data. A comprehensive social science investigation of decision makers' interests in nuclear fuel cycles, and speci cally their interests in nuclear fuel cycle simulators (NFCSs) as tools for understanding nuclear fuel cycle options, was conducted. This included document review and analysis, stakeholder interviews, and a survey of decision makers. This information was used to study the role of visualization formats and features in communicating information about nuclear fuel cycles. A exible and user-friendly tool was developed for building Cyclus analysis models, featuring a drag-and-drop interface and automatic input form generation for novel archetypes. Cycic allows users to design fuel cycles from arbitrary collections of facilities for the rst time, with mechanisms that contribute to consistency within that fuel cycle. Interacting with some of the metadata capabilities introduced in the above-mentioned tools to support archetype development, Cycic also automates the generation of user input forms for novel archetypes with little to no special knowledge required by the archetype developers. Translation of the fundamental metrics of Cyclus into more interesting quantities is

  5. Closed fuel cycle with increased fuel burn-up and economy applying of thorium resources

    Science.gov (United States)

    Kulikov, G. G.; Apse, V. A.

    2017-01-01

    The possible role of existing thorium reserves in the Russian Federation on engaging thorium in being currently closed (U-Pu)-fuel cycle of nuclear power of the country is considered. The application efficiency of thermonuclear neutron sources with thorium blanket for the economical use of existing thorium reserves is demonstrated. The aim of the work is to find solutions of such major tasks as the reduction of both front-end and back-end of nuclear fuel cycle and an enhancing its protection against the uncontrolled proliferation of fissile materials by means of the smallest changes in the fuel cycle. During implementation of the work we analyzed the results obtained earlier by the authors, brought new information on the number of thorium available in the Russian Federation and made further assessments. On the basis of proposal on the inclusion of hybrid reactors with Th-blanket into the future nuclear power for the production of light uranium fraction 232+233+234U, and 231Pa, we obtained the following results: 1. The fuel cycle will shift from fissile 235U to 233U which is more attractive for thermal power reactors. 2. The light uranium fraction is the most "protected" in the uranium component of fuel and mixed with regenerated uranium will in addition become a low enriched uranium fuel, that will weaken the problem of uncontrolled proliferation of fissile materials. 3. 231Pa doping into the fuel stabilizes its multiplying properties that will allow us to implement long-term fuel residence time and eventually to increase the export potential of all nuclear power technologies. 4. The thorium reserves being near city Krasnoufimsk (Russia) are large enough for operation of large-scale nuclear power of the Russian Federation of 70 GWe capacity during more than a quarter century under assumption that thorium is loaded into blankets of hybrid TNS only. The general conclusion: the inclusion of a small number of hybrid reactors with Th-blanket into the future nuclear

  6. A Fast Numerical Method for the Calculation of the Equilibrium Isotopic Composition of a Transmutation System in an Advanced Fuel Cycle

    Directory of Open Access Journals (Sweden)

    F. Álvarez-Velarde

    2012-01-01

    Full Text Available A fast numerical method for the calculation in a zero-dimensional approach of the equilibrium isotopic composition of an iteratively used transmutation system in an advanced fuel cycle, based on the Banach fixed point theorem, is described in this paper. The method divides the fuel cycle in successive stages: fuel fabrication, storage, irradiation inside the transmutation system, cooling, reprocessing, and incorporation of the external material into the new fresh fuel. The change of the fuel isotopic composition, represented by an isotope vector, is described in a matrix formulation. The resulting matrix equations are solved using direct methods with arbitrary precision arithmetic. The method has been successfully applied to a double-strata fuel cycle with light water reactors and accelerator-driven subcritical systems. After comparison to the results of the EVOLCODE 2.0 burn-up code, the observed differences are about a few percents in the mass estimations of the main actinides.

  7. Estimating externalities of biomass fuel cycles, Report 7

    Energy Technology Data Exchange (ETDEWEB)

    Barnthouse, L.W.; Cada, G.F.; Cheng, M.-D.; Easterly, C.E.; Kroodsma, R.L.; Lee, R.; Shriner, D.S.; Tolbert, V.R.; Turner, R.S.

    1998-01-01

    This report documents the analysis of the biomass fuel cycle, in which biomass is combusted to produce electricity. The major objectives of this study were: (1) to implement the methodological concepts which were developed in the Background Document (ORNL/RFF 1992) as a means of estimating the external costs and benefits of fuel cycles, and by so doing, to demonstrate their application to the biomass fuel cycle; (2) to develop, given the time and resources, a range of estimates of marginal (i.e., the additional or incremental) damages and benefits associated with selected impact-pathways from a new wood-fired power plant, using a representative benchmark technology, at two reference sites in the US; and (3) to assess the state of the information available to support energy decision making and the estimation of externalities, and by so doing, to assist in identifying gaps in knowledge and in setting future research agendas. The demonstration of methods, modeling procedures, and use of scientific information was the most important objective of this study. It provides an illustrative example for those who will, in the future, undertake studies of actual energy options and sites. As in most studies, a more comprehensive analysis could have been completed had budget constraints not been as severe. Particularly affected were the air and water transport modeling, estimation of ecological impacts, and economic valuation. However, the most important objective of the study was to demonstrate methods, as a detailed example for future studies. Thus, having severe budget constraints was appropriate from the standpoint that these studies could also face similar constraints. Consequently, an important result of this study is an indication of what can be done in such studies, rather than the specific numerical estimates themselves.

  8. Nuclear Fuel Cycle Reasoner: PNNL FY12 Report

    Energy Technology Data Exchange (ETDEWEB)

    Hohimer, Ryan E.; Pomiak, Yekaterina G.; Neorr, Peter A.; Gastelum, Zoe N.; Strasburg, Jana D.

    2013-05-03

    Building on previous internal investments and leveraging ongoing advancements in semantic technologies, PNNL implemented a formal reasoning framework and applied it to a specific challenge in nuclear nonproliferation. The Semantic Nonproliferation Analysis Platform (SNAP) was developed as a preliminary graphical user interface to demonstrate the potential power of the underlying semantic technologies to analyze and explore facts and relationships relating to the nuclear fuel cycle (NFC). In developing this proof of concept prototype, the utility and relevancy of semantic technologies to the Office of Defense Nuclear Nonproliferation Research and Development (DNN R&D) has been better understood.

  9. Advanced Fuel Cycle Economic Tools, Algorithms, and Methodologies

    Energy Technology Data Exchange (ETDEWEB)

    David E. Shropshire

    2009-05-01

    The Advanced Fuel Cycle Initiative (AFCI) Systems Analysis supports engineering economic analyses and trade-studies, and requires a requisite reference cost basis to support adequate analysis rigor. In this regard, the AFCI program has created a reference set of economic documentation. The documentation consists of the “Advanced Fuel Cycle (AFC) Cost Basis” report (Shropshire, et al. 2007), “AFCI Economic Analysis” report, and the “AFCI Economic Tools, Algorithms, and Methodologies Report.” Together, these documents provide the reference cost basis, cost modeling basis, and methodologies needed to support AFCI economic analysis. The application of the reference cost data in the cost and econometric systems analysis models will be supported by this report. These methodologies include: the energy/environment/economic evaluation of nuclear technology penetration in the energy market—domestic and internationally—and impacts on AFCI facility deployment, uranium resource modeling to inform the front-end fuel cycle costs, facility first-of-a-kind to nth-of-a-kind learning with application to deployment of AFCI facilities, cost tradeoffs to meet nuclear non-proliferation requirements, and international nuclear facility supply/demand analysis. The economic analysis will be performed using two cost models. VISION.ECON will be used to evaluate and compare costs under dynamic conditions, consistent with the cases and analysis performed by the AFCI Systems Analysis team. Generation IV Excel Calculations of Nuclear Systems (G4-ECONS) will provide static (snapshot-in-time) cost analysis and will provide a check on the dynamic results. In future analysis, additional AFCI measures may be developed to show the value of AFCI in closing the fuel cycle. Comparisons can show AFCI in terms of reduced global proliferation (e.g., reduction in enrichment), greater sustainability through preservation of a natural resource (e.g., reduction in uranium ore depletion), value from

  10. Business cycles and the financial performance of fuel cell companies

    Energy Technology Data Exchange (ETDEWEB)

    Henriques, I.; Sadorsky, P. [York Univ., Toronto, ON (Canada). Schulich School of Business

    2005-07-01

    Fuel cells are expected to play a major role in a hydrogen powered world. They will provide power to homes, modes of transportation and appliances. Hydrogen is the most abundant element in nature, but it must be extracted in order to be usable. It can be produced from oil, natural gas and coal or from renewable sources such as biomass, thermal or nuclear reactions. Fuel cells running on hydrogen extracted from non renewable resources have an efficiency of 30 per cent, which is twice as efficient as an internal combustion engine. The greatest barrier to mass commercialization is the cost of making hydrogen-powered auto engines. Also, an infrastructure must be developed to refill hydrogen cars. One solution is to build a hydrogen highway using the existing natural gas grid to produce hydrogen and sell it at existing filling stations. The cost of building 12,000 refueling pumps in urban areas which will provide access to 70 per cent of America's population is estimated at $10 to $15 billion. This paper described the vector autoregression (VAR) model which empirically examines the relationship between financial performance of fuel cell companies and business cycles. It was used to measure how sensitive the financial performance of fuel cell companies are to changes in macroeconomic activity. A four variable VAR model was developed to examine the relationship between stock prices, oil prices and interest rates. It was shown that the stock prices of fuel cell companies are affected by shocks to technology stock prices and oil prices, with the former having a longer lasting impact. These results add to the growing literature that oil price movements are not as important as once thought. 15 refs., 3 tabs., 3 figs.

  11. Characteristics of a different fuel cycle in a PBMR-400 for burning reactor grade plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, Eben [M-TECH Industrial (Pty) Ltd., PO Box 54620, Wierdapark 0149 (South Africa)], E-mail: ejm@mtechindustrial.com; Teuchert, Eberhard [M-TECH Industrial (Pty) Ltd., PO Box 54620, Wierdapark 0149 (South Africa)

    2008-11-15

    An outstanding feature of the high temperature, gas-cooled, version of a Generation IV type reactor is its versatility in application. Apart from the capacity in high temperature gas-cooled reactors to generate electricity it could desalinate (multi-effect distillation [MED] by deploying excess heat or reverse osmosis by deploying excess electricity), produce hydrogen (by deploying excess electricity), whereas this article showcases the on-line fuelling characteristics of a pebble bed reactor concept for the incineration of reactor grade plutonium, whilst producing electricity. The VSOP-A system of codes is employed to demonstrate by calculation how the standard PBMR-400 commercial reactor design offers similar inherent safety characteristics with a Pu-Th/U advanced fuel cycle. This implies that no significant design changes are necessary to implement such a fuel cycle. Furthermore, the flexibility of the pebble fuel concept is deployed to house the fertile material in one type of pebble, whilst a second type will contain the fissile or driver material.

  12. ENVIRONMENTAL ASSESSMENT METHODOLOGY FOR THE NUCLEAR FUEL CYCLE

    Energy Technology Data Exchange (ETDEWEB)

    Brenchley, D. L.; Soldat, J. K.; McNeese, J. A.; Watson, E. C.

    1977-07-01

    This report describes the methodology for determining where environmental control technology is required for the nuclear fuel cycle. The methodology addresses routine emission of chemical and radioactive effluents, and applies to mining, milling, conversion, enrichment, fuel fabrication, reactors (LWR and BWR) and fuel reprocessing. Chemical and radioactive effluents are evaluated independently. Radioactive effluents are evaluated on the basis of maximum exposed individual dose and population dose calculations for a 1-year emission period and a 50-year commitment. Sources of radionuclides for each facility are then listed according to their relative contribution to the total calculated dose. Effluent, ambient and toxicology standards are used to evaluate the effect of chemical effluents. First, each chemical and source configuration is determined. Sources are tagged if they exceed existirrg standards. The combined effect of all chemicals is assessed for each facility. If the additive effects are unacceptable, then additional control technology is recommended. Finally, sources and their chemicals at each facility are ranked according to their relative contribution to the ambient pollution level. This ranking identifies those sources most in need of environmental control.

  13. Transitioning nuclear fuel cycles with uncertain fast reactor costs

    Energy Technology Data Exchange (ETDEWEB)

    Phathanapirom, U.B., E-mail: bphathanapirom@utexas.edu; Schneider, E.A.

    2016-06-15

    This paper applies a novel decision making methodology to a case study involving choices leading to the transition from the current once-through light water reactor fuel cycle to one relying on continuous recycle of plutonium and minor actinides in fast reactors in the face of uncertain fast reactor capital costs. Unique to this work is a multi-stage treatment of a range of plausible trajectories for the evolution of fast reactor capital costs over time, characterized by first-of-a-kind penalties as well as time- and unit-based learning. The methodology explicitly incorporates uncertainties in key parameters into the decision-making process by constructing a stochastic model and embedding uncertainties as bifurcations in the decision tree. “Hedging” strategies are found by applying a choice criterion to select courses of action which mitigate “regrets”. These regrets are calculated by evaluating the performance of all possible transition strategies for every feasible outcome of the uncertain parameter. The hedging strategies are those that preserve the most flexibility for adjusting the fuel cycle strategy in response to new information as uncertainties are resolved.

  14. Sustainable Nuclear Fuel Cycles and World Regional Issues

    Directory of Open Access Journals (Sweden)

    Aleksandra Schwenk-Ferrero

    2012-06-01

    Full Text Available In the present paper we have attempted to associate quantified impacts with a forecasted nuclear energy development in different world regions, under a range of hypotheses on the energy demand growth. It gives results in terms of availability of uranium resources, required deployment of fuel cycle facilities and reactor types. In particular, the need to achieve short doubling times with future fast reactors is investigated and quantified in specific world regions. It has been found that a crucial feature of any world scenario study is to provide not only trends for an idealized “homogeneous” description of the global world, but also trends for different regions in the world. These regions may be selected using rather simple criteria (mostly of a geographical type, in order to apply different hypotheses for energy demand growth, fuel cycle strategies and the implementation of various reactor types for the different regions. This approach was an attempt to avoid focusing on selected countries, in particular on those where no new significant energy demand growth is expected, but instead to provide trends and conclusions that account for the features of countries that will be major players in the world energy development in the future.

  15. Changing Perspectives on Nonproliferation and Nuclear Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J; Isaacs, T H

    2005-03-29

    The concepts of international control over technologies and materials in the proliferation sensitive parts of the nuclear fuel cycle, specifically those related to enrichment and reprocessing, have been the subject of many studies and initiatives over the years. For examples: the International Fissionable Material Storage proposal in President Eisenhower's Speech on Atoms for Peace, and in the Charter of the International Atomic Energy Agency (IAEA) when the organization was formed in 1957; the regional nuclear fuel cycle center centers proposed by INFCE in the 80's; and most recently and notably, proposals by Dr. ElBaradei, the Director General of IAEA to limit production and processing of nuclear weapons usable materials to facilities under multinational control; and by U.S. President George W. Bush, to limit enrichment and reprocessing to States that have already full scale, functioning plants. There are other recent proposals on this subject as well. In this paper, the similarities and differences, as well as the effectiveness and challenges in proliferation prevention of these proposals and concepts will be discussed. The intent is to articulate a ''new nuclear regime'' and to develop concrete steps to implement such regime for future nuclear energy and deployment.

  16. NMSS handbook for decommissioning fuel cycle and materials licensees

    Energy Technology Data Exchange (ETDEWEB)

    Orlando, D.A.; Hogg, R.C.; Ramsey, K.M. [and others

    1997-03-01

    The US Nuclear Regulatory Commission amended its regulations to set forth the technical and financial criteria for decommissioning licensed nuclear facilities. These regulations were further amended to establish additional recordkeeping requirements for decommissioning; to establish timeframes and schedules for the decommissioning; and to clarify that financial assurance requirements must be in place during operations and updated when licensed operations cease. Reviews of the Site Decommissioning Management Plan (SDMP) program found that, while the NRC staff was overseeing the decommissioning program at nuclear facilities in a manner that was protective of public health and safety, progress in decommissioning many sites was slow. As a result NRC determined that formal written procedures should be developed to facilitate the timely decommissioning of licensed nuclear facilities. This handbook was developed to aid NRC staff in achieving this goal. It is intended to be used as a reference document to, and in conjunction with, NRC Inspection Manual Chapter (IMC) 2605, ``Decommissioning Inspection Program for Fuel Cycle and Materials Licensees.`` The policies and procedures discussed in this handbook should be used by NRC staff overseeing the decommissioning program at licensed fuel cycle and materials sites; formerly licensed sites for which the licenses were terminated; sites involving source, special nuclear, or byproduct material subject to NRC regulation for which a license was never issued; and sites in the NRC`s SDMP program. NRC staff overseeing the decommissioning program at nuclear reactor facilities subject to regulation under 10 CFR Part 50 are not required to use the procedures discussed in this handbook.

  17. Reactor physics modelling of accident tolerant fuel for LWRs using ANSWERS codes

    Directory of Open Access Journals (Sweden)

    Lindley Benjamin A.

    2016-01-01

    Full Text Available The majority of nuclear reactors operating in the world today and similarly the majority of near-term new build reactors will be LWRs. These currently accommodate traditional Zr clad UO2/PuO2 fuel designs which have an excellent performance record for normal operation. However, the events at Fukushima culminated in significant hydrogen production and hydrogen explosions, resulting from high temperature Zr/steam interaction following core uncovering for an extended period. These events have resulted in increased emphasis towards developing more accident tolerant fuels (ATFs-clad systems, particularly for current and near-term build LWRs. R&D programmes are underway in the US and elsewhere to develop ATFs and the UK is engaging in these international programmes. Candidate advanced fuel materials include uranium nitride (UN and uranium silicide (U3Si2. Candidate cladding materials include advanced stainless steel (FeCrAl and silicon carbide. The UK has a long history in industrial fuel manufacture and fabrication for a wide range of reactor systems including LWRs. This is supported by a national infrastructure to perform experimental and theoretical R&D in fuel performance, fuel transient behaviour and reactor physics. In this paper, an analysis of the Integral Inherently Safe LWR design (I2S-LWR, a reactor concept developed by an international collaboration led by the Georgia Institute of Technology, within a US DOE Nuclear Energy University Program (NEUP Integrated Research Project (IRP is considered. The analysis is performed using the ANSWERS reactor physics code WIMS and the EDF Energy core simulator PANTHER by researchers at the University of Cambridge. The I2S-LWR is an advanced 2850 MWt integral PWR with inherent safety features. In order to enhance the safety features, the baseline fuel and cladding materials that were chosen for the I2S-LWR design are U3Si2 and advanced stainless steel respectively. In addition, the I2S-LWR design

  18. A methodology for determining the dynamic exchange of resources in nuclear fuel cycle simulation

    Energy Technology Data Exchange (ETDEWEB)

    Gidden, Matthew J., E-mail: gidden@iiasa.ac.at [International Institute for Applied Systems Analysis, Schlossplatz 1, A-2361 Laxenburg (Austria); University of Wisconsin – Madison, Department of Nuclear Engineering and Engineering Physics, Madison, WI 53706 (United States); Wilson, Paul P.H. [University of Wisconsin – Madison, Department of Nuclear Engineering and Engineering Physics, Madison, WI 53706 (United States)

    2016-12-15

    Highlights: • A novel fuel cycle simulation entity interaction mechanism is proposed. • A framework and implementation of the mechanism is described. • New facility outage and regional interaction scenario studies are described and analyzed. - Abstract: Simulation of the nuclear fuel cycle can be performed using a wide range of techniques and methodologies. Past efforts have focused on specific fuel cycles or reactor technologies. The CYCLUS fuel cycle simulator seeks to separate the design of the simulation from the fuel cycle or technologies of interest. In order to support this separation, a robust supply–demand communication and solution framework is required. Accordingly an agent-based supply-chain framework, the Dynamic Resource Exchange (DRE), has been designed implemented in CYCLUS. It supports the communication of complex resources, namely isotopic compositions of nuclear fuel, between fuel cycle facilities and their managers (e.g., institutions and regions). Instances of supply and demand are defined as an optimization problem and solved for each timestep. Importantly, the DRE allows each agent in the simulation to independently indicate preference for specific trading options in order to meet both physics requirements and satisfy constraints imposed by potential socio-political models. To display the variety of possible simulations that the DRE enables, example scenarios are formulated and described. Important features include key fuel-cycle facility outages, introduction of external recycled fuel sources (similar to the current mixed oxide (MOX) fuel fabrication facility in the United States), and nontrivial interactions between fuel cycles existing in different regions.

  19. Development of nuclear fuel cycle technologies - bases of long-term provision of fuel and environmental safety of nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Solonin, M.I.; Polyakov, A.S.; Zakharkin, B.S.; Smelov, V.S.; Nenarokomov, E.A.; Mukhin, I.V. [SSC, RF, A.A. Bochvar ALL-Russia Research Institute of Inorganic Materials, Moscow (Russian Federation)

    2000-07-01

    To-day nuclear power is one of the options, however, to-morrow it may become the main source of the energy, thus, providing for the stable economic development for the long time to come. The availability of the large-scale nuclear power in the foreseeable future is governed by not only the safe operation of nuclear power plants (NPP) but also by the environmentally safe management of spent nuclear fuel, radioactive waste conditioning and long-term storage. More emphasis is to be placed to the closing of the fuel cycle in view of substantial quantities of spent nuclear fuel arisings. The once-through fuel cycle that is cost effective at the moment cannot be considered to be environmentally safe even for the middle term since the substantial build-up of spent nuclear fuel containing thousands of tons Pu will require the resolution of the safe management problem in the nearest future and is absolutely unjustified in terms of moral ethics as a transfer of the responsibility to future generations. The minimization of radioactive waste arisings and its radioactivity is only feasible with the closed fuel cycle put into practice and some actinides and long-lived fission radionuclides burnt out. The key issues in providing the environmentally safe fuel cycle are efficient processes of producing fuel for NPP, radionuclide after-burning included, a long-term spent nuclear fuel storage and reprocessing as well as radioactive waste management. The paper deals with the problems inherent in producing fuel for NPP with a view for the closed fuel cycle. Also discussed are options of the fuel cycle, its effectiveness and environmental safety with improvements in technologies of spent nuclear fuel reprocessing and long-lived radionuclide partitioning. (authors)

  20. Assessment of stainless steel 348 fuel rod performance against literature available data using TRANSURANUS code

    Directory of Open Access Journals (Sweden)

    Giovedi Claudia

    2016-01-01

    Full Text Available Early pressurized water reactors were originally designed to operate using stainless steel as cladding material, but during their lifetime this material was replaced by zirconium-based alloys. However, after the Fukushima Daiichi accident, the problems related to the zirconium-based alloys due to the hydrogen production and explosion under severe accident brought the importance to assess different materials. In this sense, initiatives as ATF (Accident Tolerant Fuel program are considering different material as fuel cladding and, one candidate is iron-based alloy. In order to assess the fuel performance of fuel rods manufactured using iron-based alloy as cladding material, it was necessary to select a specific stainless steel (type 348 and modify properly conventional fuel performance codes developed in the last decades. Then, 348 stainless steel mechanical and physics properties were introduced in the TRANSURANUS code. The aim of this paper is to present the obtained results concerning the verification of the modified TRANSURANUS code version against data collected from the open literature, related to reactors which operated using stainless steel as cladding. Considering that some data were not available, some assumptions had to be made. Important differences related to the conventional fuel rods were taken into account. Obtained results regarding the cladding behavior are in agreement with available information. This constitutes an evidence of the modified TRANSURANUS code capabilities to perform fuel rod investigation of fuel rods manufactured using 348 stainless steel as cladding material.

  1. Development of the thermal behavior analysis code DIRAD and the fuel design procedure for LMFBR

    Science.gov (United States)

    Nakae, N.; Tanaka, K.; Nakajima, H.; Matsumoto, M.

    1992-06-01

    It is very important to increase the fuel linear heat rating for improvement of economy in LMFBR without any degradation in safety. A reduction of the design margin is helpful to achieve the high power operation. The development of a fuel design code and a design procedure is effective on the reduction of the design margin. The thermal behavior analysis code DIRAD has been developed with respect to fuel restructuring and gap conductance models. These models have been calibrated and revised using irradiation data of fresh fuel. It is, therefore, found that the code is applicable for the thermal analysis with fresh fuel. The uncertainties in fuel irradiation condition and fuel fabrication tolerance together with the uncertainty of the code prediction have major contributions to the design margin. In the current fuel design the first two uncertainties independently contribute to temperature increment. Another method which can rationally explain the effect of the uncertainties on the temperature increment is adopted here. Then, the design margin may be rationally reduced.

  2. Origin and characteristics of low-level nontransuranic waste from the nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Alexander, C.W.; Blomeke, J.O.

    1977-01-01

    Low-level nontransuranic wastes are generated in all nuclear fuel cycle operations. While the activity levels and radiotoxicities of these effluents are generally of a lower magnitude than other fuel cycle wastes, their large volumes and their appearance throughout the fuel cycle make their management a very real concern regardless of the fuel cycle option being considered. Low-level nontransuranic wastes are defined here as wastes that contain less than about 10 nCi of long-lived alpha radiation per gram and have gamma radiations low enough to require only minimal biological shielding and remote handling. Wastes from uranium mining and milling, UF/sub 6/ conversion, enrichment, fuel fabrication, and fuel reprocessing are examined with respect to their radionuclide content, volume, and chemical composition. Projections of total quantities through the end of this century are also presented. Fuel cycles based on recycling only uranium, and on recycling both uranium and plutonium, are considered.

  3. High-Level Functional and Operational Requirements for the Advanced Fuel Cycle Facilty

    Energy Technology Data Exchange (ETDEWEB)

    Charles Park

    2006-12-01

    High-Level Functional & Operational Requirements for the AFCF -This document describes the principal functional and operational requirements for the proposed Advanced Fuel Cycle Facility (AFCF). The AFCF is intended to be the world's foremost facility for nuclear fuel cycle research, technology development, and demonstration. The facility will also support the near-term mission to develop and demonstrate technology in support of fuel cycle needs identified by industry, and the long-term mission to retain and retain U.S. leadership in fuel cycle operations. The AFCF is essential to demonstrate a more proliferation-resistant fuel cycle and make long-term improvements in fuel cycle effectiveness, performance and economy.

  4. The basic features of a closed fuel cycle without fast reactors

    Science.gov (United States)

    Bobrov, E. A.; Alekseev, P. N.; Teplov, P. S.

    2017-01-01

    In this paper the basic features of a closed fuel cycle with thermal reactors are considered. The three variants of multiple Pu and U recycling in VVER reactors was investigated. The comparison of MOX and REMIX fuel approaches for closed fuel cycle with thermal reactors is presented. All variants make possible to recycle several times the total amount of Pu and U obtained from spent fuel. The reported study was funded by RFBR according to the research project № 16-38-00021

  5. Fuel Cycle Scenario Definition, Evaluation, and Trade-offs

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Gretchen E. Matthern; Jacob J. Jacobson; Christopher T. Laws; Lee C. Cadwallader; Abdellatif M. Yacout; Robert N. Hill; J. D. Smith; Andrew S. Goldmann; George Bailey

    2006-08-01

    This report aims to clarify many of the issues being discussed within the AFCI program, including Inert Matrix Fuel (IMF) versus Mixed Oxide (MOX) fuel, single-pass versus multi-pass recycling, thermal versus fast reactors, potential need for transmutation of technetium and iodine, and the value of separating cesium and strontium. It documents most of the work produced by INL, ANL, and SNL personnel under their Simulation, Evaluation, and Trade Study (SETS) work packages during FY2005 and the first half of FY2006. This report represents the first attempt to calculate a full range of metrics, covering all four AFCI program objectives - waste management, proliferation resistance, energy recovery, and systematic management/economics/safety - using a combination of "static" calculations and a system dynamic model, DYMOND. In many cases, we examine the same issue both dynamically and statically to determine the robustness of the observations. All analyses are for the U.S. reactor fleet. This is a technical report, not aimed at a policy-level audience. A wide range of options are studied to provide the technical basis for identifying the most attractive options and potential improvements. Option improvement could be vital to accomplish before the AFCI program publishes definitive cost estimates. Information from this report will be extracted and summarized in future policy-level reports. Many dynamic simulations of deploying those options are included. There are few "control knobs" for flying or piloting the fuel cycle system into the future, even though it is dark (uncertain) and controls are sluggish with slow time response: what types of reactors are built, what types of fuels are used, and the capacity of separation and fabrication plants. Piloting responsibilities are distributed among utilities, government, and regulators, compounding the challenge of making the entire system work and respond to changing circumstances. We identify four approaches that would

  6. New high burnup fuel models for NRC`s licensing audit code, FRAPCON

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Beyer, C.E.; Painter, C.L. [Pacific Northwest Laboratory, Richland, WA (United States)

    1996-03-01

    Fuel behavior models have recently been updated within the U.S. Nuclear Regulatory Commission steady-state FRAPCON code used for auditing of fuel vendor/utility-codes and analyses. These modeling updates have concentrated on providing a best estimate prediction of steady-state fuel behavior up to the maximum burnup level s of current data (60 to 65 GWd/MTU rod-average). A decade has passed since these models were last updated. Currently, some U.S. utilities and fuel vendors are requesting approval for rod-average burnups greater than 60 GWd/MTU; however, until these recent updates the NRC did not have valid fuel performance models at these higher burnup levels. Pacific Northwest Laboratory (PNL) has reviewed 15 separate effects models within the FRAPCON fuel performance code (References 1 and 2) and identified nine models that needed updating for improved prediction of fuel behavior at high burnup levels. The six separate effects models not updated were the cladding thermal properties, cladding thermal expansion, cladding creepdown, fuel specific heat, fuel thermal expansion and open gap conductance. Comparison of these models to the currently available data indicates that these models still adequately predict the data within data uncertainties. The nine models identified as needing improvement for predicting high-burnup behavior are fission gas release (FGR), fuel thermal conductivity (accounting for both high burnup effects and burnable poison additions), fuel swelling, fuel relocation, radial power distribution, fuel-cladding contact gap conductance, cladding corrosion, cladding mechanical properties and cladding axial growth. Each of the updated models will be described in the following sections and the model predictions will be compared to currently available high burnup data.

  7. Mathematical Description of Wafer-1, a Three-Dimensional Code for LWR Fuel Performance Analysis

    DEFF Research Database (Denmark)

    Kjær-Pedersen, Niels

    1975-01-01

    This article describes in detail the mathematical formulation used in the WAFER-1 code, which is presently used for three-dimensional analysis of LWR fuel pin performance. The code aims at a prediction of the local stress-strain history in the cladding, especially with regard to the ridging pheno...

  8. Review of application code and standards for mechanical and piping design of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. HANARO fuel test loop was designed in accordance with the same code and standards of nuclear power plant because HANARO FTL will be operated the high pressure and temperature same as nuclear power plant operation conditions. The objective of this study is to confirm the propriety of application code and standards for mechanical and piping of HANARO fuel test loop and to decide the technical specification of FTL systems. (author). 18 refs., 8 tabs., 6 figs.

  9. Reactivity Insertion Accident (RIA) Capability Status in the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Richard L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Folsom, Charles Pearson [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pastore, Giovanni [Idaho National Lab. (INL), Idaho Falls, ID (United States); Veeraraghavan, Swetha [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-05-01

    One of the Challenge Problems being considered within CASL relates to modelling and simulation of Light Water Reactor LWR) fuel under Reactivity Insertion Accident (RIA) conditions. BISON is the fuel performance code used within CASL for LWR fuel under both normal operating and accident conditions, and thus must be capable of addressing the RIA challenge problem. This report outlines required BISON capabilities for RIAs and describes the current status of the code. Information on recent accident capability enhancements, application of BISON to a RIA benchmark exercise, and plans for validation to RIA behavior are included.

  10. On the development of LWR fuel analysis code (1). Analysis of the FEMAXI code and proposal of a new model

    Energy Technology Data Exchange (ETDEWEB)

    Lemehov, Sergei; Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-01-01

    This report summarizes the review on the modeling features of FEMAXI code and proposal of a new theoretical equation model of clad creep on the basis of irradiation-induced microstructure change. It was pointed out that plutonium build-up in fuel matrix and non-uniform radial power profile at high burn-up affect significantly fuel behavior through the interconnected effects with such phenomena as clad irradiation-induced creep, fission gas release, fuel thermal conductivity degradation, rim porous band formation and associated fuel swelling. Therefore, these combined effects should be properly incorporated into the models of the FEMAXI code so that the code can carry out numerical analysis at the level of accuracy and elaboration that modern experimental data obtained in test reactors have. Also, the proposed new mechanistic clad creep model has a general formalism which allows the model to be flexibly applied for clad behavior analysis under normal operation conditions and power transients as well for Zr-based clad materials by the use of established out-of-pile mechanical properties. The model has been tested against experimental data, while further verification is needed with specific emphasis on power ramps and transients. (author)

  11. Tritium fuel cycle in ITER and DEMO: Issues in handling large amount of fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tanabe, T., E-mail: tanabe@nucl.kyushu-u.ac.jp [Kyushu University, Interdisciplinary Graduate School of Engineering and Sciences, Hakozaki 6-10-1, Higashi-ku, Fukuoka 812-8581 (Japan)

    2013-07-15

    Since tritium resources are very limited, economical conversion of fusion energy to electricity and/or heat must have an enough margin to keep self-fuel-sufficiency and simultaneously ensure tritium safety. To realize this, tritium inventory in reactor systems, which use ∼10{sup 17} Bq (a few kg) of tritium, should be kept as small as possible, under strict accountancy or regulation of a few tens Bq (∼pg). In present tokamaks, however, hydrogen retention rate is significantly large, i.e. more than 5% of fueled hydrogen is continuously piled up in their vacuum vessels, which must not be allowed in a reactor. Moreover, both fuelling and burning efficiencies are very poor, only a few % or less. Accordingly, huge amount of tritium fuel must be recycled. In the paper, after the introduction of tritium as a hydrogen radioisotope, important issues to establish safe and economical tritium fuel cycle for a fusion reactor will be summarized considering the handling of large amounts of tritium, i.e. confinement, leakage, contamination, permeation, regulation and tritium accountancy.

  12. Convolutional cylinder-type block-circulant cycle codes

    Directory of Open Access Journals (Sweden)

    Mohammad Gholami

    2013-06-01

    Full Text Available In this paper, we consider a class of column-weight two quasi-cyclic low-density paritycheck codes in which the girth can be large enough, as an arbitrary multiple of 8. Then we devote a convolutional form to these codes, such that their generator matrix can be obtained by elementary row and column operations on the parity-check matrix. Finally, we show that the free distance of the convolutional codes is equal to the minimum distance of their block counterparts.

  13. National briefing summaries: Nuclear fuel cycle and waste management

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Bradley, D.J.; Fletcher, J.F.; Konzek, G.J.; Lakey, L.T.; Mitchell, S.J.; Molton, P.M.; Nightingale, R.E.

    1991-04-01

    Since 1976, the International Program Support Office (IPSO) at the Pacific Northwest Laboratory (PNL) has collected and compiled publicly available information concerning foreign and international radioactive waste management programs. This National Briefing Summaries is a printout of an electronic database that has been compiled and is maintained by the IPSO staff. The database contains current information concerning the radioactive waste management programs (with supporting information on nuclear power and the nuclear fuel cycle) of most of the nations (except eastern European countries) that now have or are contemplating nuclear power, and of the multinational agencies that are active in radioactive waste management. Information in this document is included for three additional countries (China, Mexico, and USSR) compared to the prior issue. The database and this document were developed in response to needs of the US Department of Energy.

  14. National briefing summaries: Nuclear fuel cycle and waste management

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Lakey, L.T.; Silviera, D.J.

    1988-12-01

    The National Briefing Summaries is a compilation of publicly available information concerning the nuclear fuel cycle and radioactive waste management strategies and programs of 21 nations, including the United States and three international agencies that have publicized their activities in this field. It presents available highlight information with references that may be used by the reader for additional information. The information in this document is compiled primarily for use by the US Department of Energy and other US federal agencies and their contractors to provide summary information on radioactive waste management activities in other countries. This document provides an awareness to managers and technical staff of what is occurring in other countries with regard to strategies, activities, and facilities. The information may be useful in program planning to improve and benefit United States' programs through foreign information exchange. Benefits to foreign exchange may be derived through a number of exchange activities.

  15. Future nuclear fuel cycles: Prospect and challenges for actinide recycling

    Science.gov (United States)

    Warin, Dominique

    2010-03-01

    The global energy context pleads in favour of a sustainable development of nuclear energy since the demand for energy will likely increase, whereas resources will tend to get scarcer and the prospect of global warming will drive down the consumption of fossil fuel. In this context, nuclear power has the worldwide potential to curtail the dependence on fossil fuels and thereby to reduce the amount of greenhouse gas emissions while promoting energy independence. How we deal with nuclear radioactive waste is crucial in this context. In France, the public's concern regarding the long-term waste management made the French Governments to prepare and pass the 1991 and 2006 Acts, requesting in particular the study of applicable solutions for still minimizing the quantity and the hazardousness of final waste. This necessitates High Active Long Life element (such as the Minor Actinides MA) recycling, since the results of fuel cycle R&D could significantly change the challenges for the storage of nuclear waste. HALL recycling can reduce the heat load and the half-life of most of the waste to be buried to a couple of hundred years, overcoming the concerns of the public related to the long-life of the waste and thus aiding the "burying approach" in securing a "broadly agreed political consensus" of waste disposal in a geological repository. This paper presents an overview of the recent R and D results obtained at the CEA Atalante facility on innovative actinide partitioning hydrometallurgical processes. For americium and curium partitioning, these results concern improvements and possible simplifications of the Diamex-Sanex process, whose technical feasibility was already demonstrated in 2005. Results on the first tests of the Ganex process (grouped actinide separation for homogeneous recycling) are also discussed. In the coming years, next steps will involve both better in-depth understanding of the basis of these actinide partitioning processes and, for the new promising

  16. The FIT 2.0 Model - Fuel-cycle Integration and Tradeoffs

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Nick R. Soelberg; Layne F. Pincock; Eric L. Shaber; Gregory M Teske

    2011-06-01

    All mass streams from fuel separation and fabrication are products that must meet some set of product criteria – fuel feedstock impurity limits, waste acceptance criteria (WAC), material storage (if any), or recycle material purity requirements such as zirconium for cladding or lanthanides for industrial use. These must be considered in a systematic and comprehensive way. The FIT model and the “system losses study” team that developed it [Shropshire2009, Piet2010b] are steps by the Fuel Cycle Technology program toward an analysis that accounts for the requirements and capabilities of each fuel cycle component, as well as major material flows within an integrated fuel cycle. This will help the program identify near-term R&D needs and set longer-term goals. This report describes FIT 2, an update of the original FIT model.[Piet2010c] FIT is a method to analyze different fuel cycles; in particular, to determine how changes in one part of a fuel cycle (say, fuel burnup, cooling, or separation efficiencies) chemically affect other parts of the fuel cycle. FIT provides the following: Rough estimate of physics and mass balance feasibility of combinations of technologies. If feasibility is an issue, it provides an estimate of how performance would have to change to achieve feasibility. Estimate of impurities in fuel and impurities in waste as function of separation performance, fuel fabrication, reactor, uranium source, etc.

  17. The FIT 2.0 Model - Fuel-cycle Integration and Tradeoffs

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Nick R. Soelberg; Layne F. Pincock; Eric L. Shaber; Gregory M Teske

    2011-06-01

    All mass streams from fuel separation and fabrication are products that must meet some set of product criteria – fuel feedstock impurity limits, waste acceptance criteria (WAC), material storage (if any), or recycle material purity requirements such as zirconium for cladding or lanthanides for industrial use. These must be considered in a systematic and comprehensive way. The FIT model and the “system losses study” team that developed it [Shropshire2009, Piet2010b] are steps by the Fuel Cycle Technology program toward an analysis that accounts for the requirements and capabilities of each fuel cycle component, as well as major material flows within an integrated fuel cycle. This will help the program identify near-term R&D needs and set longer-term goals. This report describes FIT 2, an update of the original FIT model.[Piet2010c] FIT is a method to analyze different fuel cycles; in particular, to determine how changes in one part of a fuel cycle (say, fuel burnup, cooling, or separation efficiencies) chemically affect other parts of the fuel cycle. FIT provides the following: Rough estimate of physics and mass balance feasibility of combinations of technologies. If feasibility is an issue, it provides an estimate of how performance would have to change to achieve feasibility. Estimate of impurities in fuel and impurities in waste as function of separation performance, fuel fabrication, reactor, uranium source, etc.

  18. Long-term global nuclear energy and fuel cycle strategies

    Energy Technology Data Exchange (ETDEWEB)

    Krakowski, R.A. [Los Alamos National Lab., NM (United States). Technology and Safety Assessment Div.

    1997-09-24

    The Global Nuclear Vision Project is examining, using scenario building techniques, a range of long-term nuclear energy futures. The exploration and assessment of optimal nuclear fuel-cycle and material strategies is an essential element of the study. To this end, an established global E{sup 3} (energy/economics/environmental) model has been adopted and modified with a simplified, but comprehensive and multi-regional, nuclear energy module. Consistent nuclear energy scenarios are constructed using this multi-regional E{sup 3} model, wherein future demands for nuclear power are projected in price competition with other energy sources under a wide range of long-term demographic (population, workforce size and productivity), economic (price-, population-, and income-determined demand for energy services, price- and population-modified GNP, resource depletion, world-market fossil energy prices), policy (taxes, tariffs, sanctions), and top-level technological (energy intensity and end-use efficiency improvements) drivers. Using the framework provided by the global E{sup 3} model, the impacts of both external and internal drivers are investigated. The ability to connect external and internal drivers through this modeling framework allows the study of impacts and tradeoffs between fossil- versus nuclear-fuel burning, that includes interactions between cost, environmental, proliferation, resource, and policy issues.

  19. Concept of DT fuel cycle for a fusion neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Anan' ev, S.; Spitsyn, A.V.; Kuteev, B.V.; Cherkez, D.I. [NRC Kurchatov Institute, Moscow (Russian Federation); Shirnin, P.N.; Kazakovsky, N.T. [FSUE RFNC - VNIIEF, Sarov (Russian Federation)

    2015-03-15

    A concept of DT-fusion neutron source (FNS) with the neutron yield higher than 10{sup 18} neutrons per second is under design in Russia. Such a FNS is of interest for many applications: 1) basic and applied research (neutron scattering, etc); 2) testing the structural materials for fusion reactors; 3) control of sub-critical nuclear systems and 4) nuclear waste processing (including transmutation of minor actinides). This paper describes the fuel cycle concept of a compact fusion neutron source based on a small spherical tokamak (FNS-ST) with a MW range of DT fusion power and considers the key physics issues of this device. The major and minor radii are ∼0.5 and ∼0.3 m, magnetic field ∼1.5 T, heating power less than 15 MW and plasma current 1-2 MA. The system provides the fuel mixture with equal fractions of D and T (D:T = 1:1) for all FNS technology systems. (authors)

  20. Perodua Myvi engine fuel consumption map and fuel economy vehicle simulation on the drive cycles based on Malaysian roads

    Directory of Open Access Journals (Sweden)

    Ramdan Muhammad Iftishah

    2017-01-01

    Full Text Available This paper presents the fuel consumption engine map for a 1.3L Perodua Myvi passenger car. The engine dynamometer and the engine throttle are controlled, to create the operating conditions for the engine map. Interpolation work is done in MATLAB, to create a 3D fuel consumption engine map. The engine map is used in a fuel-economy estimation simulation, using the city and the highway drive cycles based on Malaysian roads. The fuel economy values generated from the simulations are similar to experimental fuel consumption results.

  1. Inclusion of models to describe severe accident conditions in the fuel simulation code DIONISIO

    Energy Technology Data Exchange (ETDEWEB)

    Lemes, Martín; Soba, Alejandro [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Daverio, Hernando [Gerencia Reactores y Centrales Nucleares, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Denis, Alicia [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina)

    2017-04-15

    The simulation of fuel rod behavior is a complex task that demands not only accurate models to describe the numerous phenomena occurring in the pellet, cladding and internal rod atmosphere but also an adequate interconnection between them. In the last years several models have been incorporated to the DIONISIO code with the purpose of increasing its precision and reliability. After the regrettable events at Fukushima, the need for codes capable of simulating nuclear fuels under accident conditions has come forth. Heat removal occurs in a quite different way than during normal operation and this fact determines a completely new set of conditions for the fuel materials. A detailed description of the different regimes the coolant may exhibit in such a wide variety of scenarios requires a thermal-hydraulic formulation not suitable to be included in a fuel performance code. Moreover, there exist a number of reliable and famous codes that perform this task. Nevertheless, and keeping in mind the purpose of building a code focused on the fuel behavior, a subroutine was developed for the DIONISIO code that performs a simplified analysis of the coolant in a PWR, restricted to the more representative situations and provides to the fuel simulation the boundary conditions necessary to reproduce accidental situations. In the present work this subroutine is described and the results of different comparisons with experimental data and with thermal-hydraulic codes are offered. It is verified that, in spite of its comparative simplicity, the predictions of this module of DIONISIO do not differ significantly from those of the specific, complex codes.

  2. Nuclear Fuel Cycle Analysis by Integrated AHP and TOPSIS Method Using an Equilibrium Model

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, S. R. [University of Science and Technology, Daejeon (Korea, Republic of); Choi, S. Y. [UNIST, Ulju (Korea, Republic of); Koc, W. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Determining whether to break away from domestic conflict surrounding nuclear power and step forward for public consensus can be identified by transparent policy making considering public acceptability. In this context, deriving the best suitable nuclear fuel cycle for Korea is the key task in current situation. Assessing nuclear fuel cycle is a multicriteria decision making problem dealing with multiple interconnected issues on efficiently using natural uranium resources, securing an environment friendliness to deal with waste, obtaining the public acceptance, ensuring peaceful uses of nuclear energy, maintaining economic competitiveness compared to other electricity sources, and assessing technical feasibility of advanced nuclear energy systems. This paper performed the integrated AHP and TOPSIS analysis on three nuclear fuel cycle options against 5 different criteria including U utilization, waste management, material attractiveness, economics, and technical feasibility. The fuel cycle options analyzed in this paper are three different fuel cycle options as follows: PWR-Once through cycle(PWR-OT), PWR-MOX cycle, Pyro- SFR cycle. These fuel cycles are most likely to be adopted in the foreseeable future. Analytic Hierarchy Process (AHP) and TOPSIS (Technique for Order of Preference by Similarity to Ideal Solution). The analyzed nuclear fuel cycle options include the once-through cycle, the PWR-MOX recycle, and the Pyro-SFR recycle.

  3. Study on the fuel cycle cost of gas turbine high temperature reactor (GTHTR300). Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Takei, Masanobu; Katanishi, Shoji; Nakata, Tetsuo; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Oda, Takefumi; Izumiya, Toru [Nuclear Fuel Industries, Ltd., Tokyo (Japan)

    2002-11-01

    In the basic design of gas turbine high temperature reactor (GTHTR300), reduction of the fuel cycle cost has a large benefit of improving overall plant economy. Then, fuel cycle cost was evaluated for GTHTR300. First, of fuel fabrication for high-temperature gas cooled reactor, since there was no actual experience with a commercial scale, a preliminary design for a fuel fabrication plant with annual processing of 7.7 ton-U sufficient four GTHTR300 was performed, and fuel fabrication cost was evaluated. Second, fuel cycle cost was evaluated based on the equilibrium cycle of GTHTR300. The factors which were considered in this cost evaluation include uranium price, conversion, enrichment, fabrication, storage of spent fuel, reprocessing, and waste disposal. The fuel cycle cost of GTHTR300 was estimated at about 1.07 yen/kWh. If the back-end cost of reprocessing and waste disposal is included and assumed to be nearly equivalent to LWR, the fuel cycle cost of GTHTR300 was estimated to be about 1.31 yen/kWh. Furthermore, the effects on fuel fabrication cost by such of fuel specification parameters as enrichment, the number of fuel types, and the layer thickness were considered. Even if the enrichment varies from 10 to 20%, the number of fuel types change from 1 to 4, the 1st layer thickness of fuel changes by 30 {mu}m, or the 2nd layer to the 4th layer thickness of fuel changes by 10 {mu}m, the impact on fuel fabrication cost was evaluated to be negligible. (author)

  4. Descriptions of reference LWR facilities for analysis of nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Kabele, T.J.

    1979-09-01

    To contribute to the Department of Energy's identification of needs for improved environmental controls in nuclear fuel cycles, a study was made of a light water reactor system. A reference LWR fuel cycle was defined, and each step in this cycle was characterized by facility description and mainline and effluent treatment process performance. The reference fuel cycle uses fresh uranium in light water reactors. Final treatment and ultimate disposition of waste from the fuel cycle steps were not included, and the waste is assumed to be disposed of by approved but currently undefined means. The characterization of the reference fuel cycle system is intended as basic information for further evaluation of alternative effluent control systems.

  5. Descriptions of reference LWR facilities for analysis of nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Kabele, T.J.

    1979-09-01

    To contribute to the Department of Energy's identification of needs for improved environmental controls in nuclear fuel cycles, a study was made of a light water reactor system. A reference LWR fuel cycle was defined, and each step in this cycle was characterized by facility description and mainline and effluent treatment process performance. The reference fuel cycle uses fresh uranium in light water reactors. Final treatment and ultimate disposition of waste from the fuel cycle steps were not included, and the waste is assumed to be disposed of by approved but currently undefined means. The characterization of the reference fuel cycle system is intended as basic information for further evaluation of alternative effluent control systems.

  6. Sensitivity Analysis of FEAST-Metal Fuel Performance Code: Initial Results

    Energy Technology Data Exchange (ETDEWEB)

    Edelmann, Paul Guy [Los Alamos National Laboratory; Williams, Brian J. [Los Alamos National Laboratory; Unal, Cetin [Los Alamos National Laboratory; Yacout, Abdellatif [Argonne National Laboratories

    2012-06-27

    This memo documents the completion of the LANL milestone, M3FT-12LA0202041, describing methodologies and initial results using FEAST-Metal. The FEAST-Metal code calculations for this work are being conducted at LANL in support of on-going activities related to sensitivity analysis of fuel performance codes. The objective is to identify important macroscopic parameters of interest to modeling and simulation of metallic fuel performance. This report summarizes our preliminary results for the sensitivity analysis using 6 calibration datasets for metallic fuel developed at ANL for EBR-II experiments. Sensitivity ranking methodology was deployed to narrow down the selected parameters for the current study. There are approximately 84 calibration parameters in the FEAST-Metal code, of which 32 were ultimately used in Phase II of this study. Preliminary results of this sensitivity analysis led to the following ranking of FEAST models for future calibration and improvements: fuel conductivity, fission gas transport/release, fuel creep, and precipitation kinetics. More validation data is needed to validate calibrated parameter distributions for future uncertainty quantification studies with FEAST-Metal. Results of this study also served to point out some code deficiencies and possible errors, and these are being investigated in order to determine root causes and to improve upon the existing code models.

  7. Fuels and cycles for the fourth generation systems; Combustibles et cycles pour les systemes de 4. generation

    Energy Technology Data Exchange (ETDEWEB)

    Brossard, Ph. [CEA Saclay, Dir. du Developpement et de l' Innovation Nucleares (DEN/DDIN), 91 - Gif Sur Yvette (France)

    2003-07-01

    This paper presented during the Cea Seminar on the nuclear systems of the future deals with the fuels and cycles of the fourth generation systems. Four goal areas have been defined for these systems: sustainable for the environment, economical for the investment, safety and reliability, a better protection against the proliferation. The different fuels and cycles of the research programs are detailed. (A.L.B.)

  8. THE MISSION AND ACCOMPLISHMENTS FROM DOE’S FUEL CYCLE RESEARCH AND DEVELOPMENT (FCRD) ADVANCED FUELS CAMPAIGN

    Energy Technology Data Exchange (ETDEWEB)

    J. Carmack; L. Braase; F. Goldner

    2015-09-01

    The mission of the Advanced Fuels Campaign (AFC) is to perform Research, Development, and Demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors, enhance proliferation resistance of nuclear fuel, effectively utilize nuclear energy resources, and address the longer-term waste management challenges. This includes development of a state of the art Research and Development (R&D) infrastructure to support the use of a “goal oriented science based approach.” AFC uses a “goal oriented, science based approach” aimed at a fundamental understanding of fuel and cladding fabrication methods and performance under irradiation, enabling the pursuit of multiple fuel forms for future fuel cycle options. This approach includes fundamental experiments, theory, and advanced modeling and simulation. One of the most challenging aspects of AFC is the management, integration, and coordination of major R&D activities across multiple organizations. AFC interfaces and collaborates with Fuel Cycle Technologies (FCT) campaigns, universities, industry, various DOE programs and laboratories, federal agencies (e.g., Nuclear Regulatory Commission [NRC]), and international organizations. Key challenges are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Challenged with the research and development of fuels for two different reactor technology platforms, AFC targeted transmutation fuel development and focused ceramic fuel development for Advanced LWR Fuels.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Thermal cycling effect in U-10Mo/Zry-4 monolithic nuclear fuel

    Science.gov (United States)

    Lopes, Denise A.; Zimmermann, Angelo J. O.; Silva, Selma L.; Piqueira, J. R. C.

    2016-05-01

    Uranium alloys in a monolithic form have been considered attractive candidates for high density nuclear fuel. However, this high-density fissile material configuration keeps the volume permitted for the retention of fission products at a minimum. Additionally, the monolithic nuclear fuel has a peculiar configuration, whereby the fuel is in direct contact with the cladding. How this fuel configuration will retain fission products and how this will affect its integrity under various physical conditions - such as thermal cycling - are some of the technological problems for this new fuel. In this paper, the effect of out-of-pile thermal cycling is studied for a monolithic fuel plate produced by a hot co-rolling method using U-10Mo (wt %) as the fuel alloy and Zircaloy-4 as the cladding material. After performing 10 thermal cycles from 25 to 400 °C at a rate of 1 °C/min (∼125 h), the fuel alloy presented several fractures that were observed to occur in the last three cycles. These cracks nucleated approximately in the center of the fuel alloy and crossed the interdiffusion zone initiating an internal crack in the cladding. The results suggest that the origin of these fractures is the thermal fatigue of the U-10Mo alloy caused due to the combination of two factors: (i) the high difference in the thermal expansion coefficient of the fuel and of the cladding material, and (ii) the bound condition of fuel/cladding materials in this fuel element configuration.

  11. Life-Cycle Assessment of Cookstove Fuels in India and China

    Science.gov (United States)

    A life cycle assessment (LCA) was conducted to compare the environmental footprint of current and possible fuels used for cooking within China and India. Current fuel mix profiles are compared to scenarios of projected differences in and/or cleaner cooking fuels. Results are repo...

  12. To Recycle or Not to Recycle? An Intergenerational Approach to Nuclear Fuel Cycles

    NARCIS (Netherlands)

    Taebi, B.; Kloosterman, J.L.

    2007-01-01

    AbstractThis paper approaches the choice between the open and closed nuclear fuel cycles as a matter of intergenerational justice, by revealing the value conflicts in the production of nuclear energy. The closed fuel cycle improve sustainability in terms of the supply certainty of uranium and involv

  13. 78 FR 45983 - Acceptability of Corrective Action Programs for Fuel Cycle Facilities

    Science.gov (United States)

    2013-07-30

    ...The U.S. Nuclear Regulatory Commission (NRC) is withdrawing draft NUREG-2154, ``Acceptability of Corrective Action Programs for Fuel Cycle Facilities,'' based on receipt and review of public comments. The draft NUREG provided guidance to NRC staff on how to determine whether a submittal for a Corrective Action Program (CAP), voluntarily submitted by fuel cycle facility licensees, was......

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

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

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

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

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

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

  16. Completion of Population of and Quality Assurance on the Nuclear Fuel Cycle Options Catalog.

    Energy Technology Data Exchange (ETDEWEB)

    Price, Laura L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Barela, Amanda Crystal [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Walkow, Walter M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Schetnan, Richard Reed [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Arnold, Matthew Brian [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-12-01

    An Evaluation and Screening team supporting the Fuel Cycle Technologies Program Office of the United States Department of Energy, Office of Nuclear Energy is conducting an evaluation and screening of a comprehensive set of fuel cycle options. These options have been assigned to one of 40 evaluation groups, each of which has a representative fuel cycle option [Todosow 2013]. A Fuel Cycle Data Package System Datasheet has been prepared for each representative fuel cycle option to ensure that the technical information used in the evaluation is high-quality and traceable [Kim, et al., 2013]. The information contained in the Fuel Cycle Data Packages has been entered into the Nuclear Fuel Cycle Options Catalog at Sandia National Laboratories so that it is accessible by the evaluation and screening team and other interested parties. In addition, an independent team at Savannah River National Laboratory has verified that the information has been entered into the catalog correctly. This report documents that the 40 representative fuel cycle options have been entered into the Catalog, and that the data entered into the catalog for the 40 representative options has been entered correctly.

  17. On feasibility of a closed nuclear power fuel cycle with minimum radioactivity

    Science.gov (United States)

    Andrianova, E. A.; Davidenko, V. D.; Tsibulskiy, V. F.

    2015-12-01

    Practical implementation of a closed nuclear fuel cycle implies solution of two main tasks. The first task is creation of environmentally acceptable operating conditions of the nuclear fuel cycle considering, first of all, high radioactivity of the involved materials. The second task is creation of effective and economically appropriate conditions of involving fertile isotopes in the fuel cycle. Creation of technologies for management of the high-level radioactivity of spent fuel reliable in terms of radiological protection seems to be the hardest problem.

  18. Advanced fuel cycle cost estimation model and its cost estimation results for three nuclear fuel cycles using a dynamic model in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sungki, E-mail: sgkim1@kaeri.re.kr [Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Ko, Wonil [Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Youn, Saerom; Gao, Ruxing [University of Science and Technology, 217 Gajungro, Yuseong-gu, Daejeon 305-350 (Korea, Republic of); Bang, Sungsig, E-mail: ssbang@kaist.ac.kr [Korea Advanced Institute of Science and Technology, Department of Business and Technology Management, 291 Deahak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2015-11-15

    Highlights: • The nuclear fuel cycle cost using a new cost estimation model was analyzed. • The material flows of three nuclear fuel cycle options were calculated. • The generation cost of once-through was estimated to be 66.88 mills/kW h. • The generation cost of pyro-SFR recycling was estimated to be 78.06 mills/kW h. • The reactor cost was identified as the main cost driver of pyro-SFR recycling. - Abstract: The present study analyzes advanced nuclear fuel cycle cost estimation models such as the different discount rate model and its cost estimation results. To do so, an analysis of the nuclear fuel cycle cost of three options (direct disposal (once through), PWR–MOX (Mixed OXide fuel), and Pyro-SFR (Sodium-cooled Fast Reactor)) from the viewpoint of economic sense, focusing on the cost estimation model, was conducted using a dynamic model. From an analysis of the fuel cycle cost estimation results, it was found that some cost gap exists between the traditional same discount rate model and the advanced different discount rate model. However, this gap does not change the priority of the nuclear fuel cycle option from the viewpoint of economics. In addition, the fuel cycle costs of OT (Once-Through) and Pyro-SFR recycling based on the most likely value using a probabilistic cost estimation except for reactor costs were calculated to be 8.75 mills/kW h and 8.30 mills/kW h, respectively. Namely, the Pyro-SFR recycling option was more economical than the direct disposal option. However, if the reactor cost is considered, the economic sense in the generation cost between the two options (direct disposal vs. Pyro-SFR recycling) can be changed because of the high reactor cost of an SFR.

  19. Economic incentives and recommended development for commercial use of high burnup fuels in the once-through LWR fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Stout, R.B.; Merckx, K.R.; Holm, J.S.

    1981-01-01

    This study calculates the reduced uranium requirements and the economic incentives for increasing the burnup of current design LWR fuels from the current range of 25 to 35 MWD/Kg to a range of 45 to 55 MWD/Kg. The changes in fuel management strategies which may be required to accommodate these high burnup fuels and longer fuel cycles are discussed. The material behavior problems which may present obstacles to achieving high burnup or to license fuel are identified and discussed. These problems are presented in terms of integral fuel response and the informational needs for commercial and licensing acceptance. Research and development programs are outlined which are aimed at achieving a licensing position and commercial acceptance of high burnup fuels.

  20. Life cycle assessment integrated with thermodynamic analysis of bio-fuel options for solid oxide fuel cells.

    Science.gov (United States)

    Lin, Jiefeng; Babbitt, Callie W; Trabold, Thomas A

    2013-01-01

    A methodology that integrates life cycle assessment (LCA) with thermodynamic analysis is developed and applied to evaluate the environmental impacts of producing biofuels from waste biomass, including biodiesel from waste cooking oil, ethanol from corn stover, and compressed natural gas from municipal solid wastes. Solid oxide fuel cell-based auxiliary power units using bio-fuel as the hydrogen precursor enable generation of auxiliary electricity for idling heavy-duty trucks. Thermodynamic analysis is applied to evaluate the fuel conversion efficiency and determine the amount of fuel feedstock needed to generate a unit of electrical power. These inputs feed into an LCA that compares energy consumption and greenhouse gas emissions of different fuel pathways. Results show that compressed natural gas from municipal solid wastes is an optimal bio-fuel option for SOFC-APU applications in New York State. However, this methodology can be regionalized within the U.S. or internationally to account for different fuel feedstock options.

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

    Energy Technology Data Exchange (ETDEWEB)

    Lindley, Benjamin A.; Parks, Geoffrey T. [University of Cambridge, Cambridge (United Kingdom); Franceschini, Fausto [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    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)

  2. Degree of Sustainability of Various Nuclear Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Brogli, R.; Krakowski, R.A. [Los Alamos National Laboratory, New Mexico (United States)

    2002-08-01

    The focus of this study is on a 'top-level' examination of the sustainability of nuclear energy in the context of the overall nuclear fuel cycle (NFC). This evaluation is conducted according to a set of established sustainability criteria that encompasses key economic (energy generation costs), environmental (resource utilization, long-term waste accumulations), and societal (nuclear-weapons proliferation risk) concerns associated with present and future NFC approaches. In this study, key NFCs are assessed according to a simplified and limited set of criteria that attempts to quantify NFC concerns related to cost, resource, waste, and proliferation. The overarching aim of this study is to examine a representative set of NFC options on a relative basis according to the adopted set of criteria to aid in the assessment and decision-making process. These criteria were then aggregated into a single, composite metric to examine the impacts of specific 'stakeholder' preferences. The study architecture is based on sets of nuclear process components. These sets are assembled around a particular nuclear reactor technology for the generation of electricity. Selections are made from the resulting sets of reactor-centric technologies and grouped to form nine central NFC scenarios. The above-described sustainability metrics are evaluated using a steady-state (equilibrium), highly aggregated model that is applied through mass and energy conservation to evaluate each NFC scenario. Six NFC scenarios examined to varying degrees are adaptations or extensions of scenarios used in a recent OECD study (OECD, 2002) of partitioning and transmutation (P and T) schemes based on accelerator-driven systems (ADS) or fast reactors (FR). Three NFC scenarios are based entirely on present-day or near-term LWR technologies. In addition to these near-term scenarios, more advanced systems considered in the original OECD study on which this model is based were retained using a

  3. JAEA key facilities for global advanced fuel cycle R and D

    Energy Technology Data Exchange (ETDEWEB)

    Nomura, Shigeo; Yamamoto, Ryuichi [Nuclear Fuel Cycle Engineering Labos, JAEA, 4-33 Tokai-mura, Ibaraki, 319-1194 (Japan)

    2008-07-01

    Advanced fuel cycle will be realized with the mid and long term R and D during the long-term transition period from LWR cycle to advanced reactor fuel cycle. Most of JAEA facilities have been utilized to establish the current LWR and FBR (Fast Breeder Reactor) fuel cycle by implementing evolutionary R and D. An assessment of today's state experimental facilities concerning the following research issues: reprocessing, Mox fuel fabrication, irradiation and post-irradiation examination, waste management and nuclear data measurement, is made. The revolutionary R and D requests new issues to be studied: the TRU multi-recycling, minor actinide recycling, the assessment of proliferation resistance and the assessment of cost reduction. To implement the revolutionary R and D for advanced fuel cycle, however, these facilities should be refurbished to install new machines and process equipment to provide more flexible testing parameters.

  4. Solar fuel processing efficiency for ceria redox cycling using alternative oxygen partial pressure reduction methods

    OpenAIRE

    Lin, Meng; Haussener, Sophia

    2015-01-01

    Solar-driven non-stoichiometric thermochemical redox cycling of ceria for the conversion of solar energy into fuels shows promise in achieving high solar-to-fuel efficiency. This efficiency is significantly affected by the operating conditions, e.g. redox temperatures, reduction and oxidation pressures, solar irradiation concentration, or heat recovery effectiveness. We present a thermodynamic analysis of five redox cycle designs to investigate the effects of working conditions on the fuel pr...

  5. Summary of non-US national and international fuel cycle and radioactive waste management programs 1982

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, K.M.; Kelman, J.A.

    1982-08-01

    Brief program overviews of fuel cycle, spent fuel, and waste management activities in the following countries are provided: Argentina, Australia, Austria, Belgium, Brazil, Canada, China, Denmark, Finland, France, German Federal Republic, India, Italy, Japan, Republic of Korea, Mexico, Netherlands, Pakistan, South Africa, Spain, Sweden, Switzerland, Taiwan, USSR, and the United Kingdom. International nonproliferation activities, multilateral agreements and projects, and the international agencies specifically involved in the nuclear fuel cycle are also described.

  6. Revised uranium--plutonium cycle PWR and BWR models for the ORIGEN computer code

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A. G.; Bjerke, M. A.; Morrison, G. W.; Petrie, L. M.

    1978-09-01

    Reactor physics calculations and literature searches have been conducted, leading to the creation of revised enriched-uranium and enriched-uranium/mixed-oxide-fueled PWR and BWR reactor models for the ORIGEN computer code. These ORIGEN reactor models are based on cross sections that have been taken directly from the reactor physics codes and eliminate the need to make adjustments in uncorrected cross sections in order to obtain correct depletion results. Revised values of the ORIGEN flux parameters THERM, RES, and FAST were calculated along with new parameters related to the activation of fuel-assembly structural materials not located in the active fuel zone. Recommended fuel and structural material masses and compositions are presented. A summary of the new ORIGEN reactor models is given.

  7. Transition Analysis of Promising U.S. Future Fuel Cycles Using ORION

    Energy Technology Data Exchange (ETDEWEB)

    Sunny, Eva E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peterson, Joshua L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gehin, Jess C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gregg, Robert [Preston Laboratory, Springfields Works (United Kingdom)

    2015-01-01

    The US Department of Energy Office of Fuel Cycle Technologies performed an evaluation and screening (E&S) study of nuclear fuel cycle options to help prioritize future research and development decisions. Previous work for this E&S study focused on establishing equilibrium conditions for analysis examples of 40 nuclear fuel cycle evaluation groups (EGs) and evaluating their performance according to a set of 22 standardized metrics. Following the E&S study, additional studies are being conducted to assess transitioning from the current US fuel cycle to future fuel cycle options identified by the E&S study as being most promising. These studies help inform decisions on how to effectively achieve full transition, estimate the length of time needed to undergo transition from the current fuel cycle, and evaluate performance of nuclear systems and facilities in place during the transition. These studies also help identify any barriers to achieve transition. Oak Ridge National Laboratory (ORNL) Fuel Cycle Options Campaign team used ORION to analyze the transition pathway from the existing US nuclear fuel cycle—the once-through use of low-enriched-uranium (LEU) fuel in thermal-spectrum light water reactors (LWRs) —to a new fuel cycle with continuous recycling of plutonium and uranium in sodium fast reactors (SFRs). This paper discusses the analysis of the transition from an LWR to an SFR fleet using ORION, highlights the role of lifetime extensions of existing LWRs to aid transition, and discusses how a slight delay in SFR deployment can actually reduce the time to achieve an equilibrium fuel cycle.

  8. Uranium in the Nuclear Fuel Cycle: Creation of Plutonium (Invited)

    Science.gov (United States)

    Ewing, R. C.

    2009-12-01

    One of the important properties of uranium is that it can be used to “breed” higher actinides, particularly plutonium. During the past sixty years, more than 1,800 metric tonnes of Pu, and substantial quantities of the “minor” actinides, such as Np, Am and Cm, have been generated in nuclear reactors - a permanent record of nuclear power. Some of these transuranium elements can be a source of energy in fission reactions (e.g., 239Pu), a source of fissile material for nuclear weapons (e.g., 239Pu and 237Np), and of environmental concern because of their long-half lives and radiotoxicity (e.g., 239Pu and 237Np). In fact, the new strategies of the Advance Fuel Cycle Initiative (AFCI) are, in part, motivated by an effort to mitigate some of the challenges of the disposal of these long-lived actinides. There are two basic strategies for the disposition of these heavy elements: 1.) to “burn” or transmute the actinides using nuclear reactors or accelerators; 2.) to “sequester” the actinides in chemically durable, radiation-resistant materials that are suitable for geologic disposal. There has been substantial interest in the use of actinide-bearing minerals, such as zircon or isometric pyrochlore, A2B2O7 (A= rare earths; B = Ti, Zr, Sn, Hf), for the immobilization of actinides, particularly plutonium, both as inert matrix fuels and nuclear waste forms. Systematic studies of rare-earth pyrochlores have led to the discovery that certain compositions (B = Zr, Hf) are stable to very high doses of alpha-decay event damage1. The radiation stability of these compositions is closely related to the structural distortions that can be accommodated for specific pyrochlore compositions and the electronic structure of the B-site cation. Recent developments in the understanding of the properties of heavy element solids have opened up new possibilities for the design of advanced nuclear fuels and waste forms.

  9. Constituent Redistribution in U-Zr Metallic Fuel Using the Advanced Fuel Performance Code BISON

    Energy Technology Data Exchange (ETDEWEB)

    Galloway, Jack D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Unal, Cetin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Matthews, Christopher [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-30

    Previous work done by Galloway, et. al. on EBR-II ternary (U-Pu-Zr) fuel constituent redistribution yielded accurate simulation data for the limited data sets of Zr redistribution. The data sets included EPMA scans of two different irradiated rods. First, T179, which was irradiated to 1.9 at% burnup, was analyzed. Second, DP16, which was irradiated to 11 at% burnup, was analyzed. One set of parameters that most accurately represented the zirconium profiles for both experiments was determined. Since the binary fuel (U-Zr) has previously been used as the driver fuel for sodium fast reactors (SFR) as well as being the likely driver fuel if a new SFR is constructed, this same process has been initiated on the binary fuel form. From limited binary EPMA scans as well as other fuel characterization techniques, it has been observed that zirconium redistribution also occurs in the binary fuel, albeit at a reduced rate compared to observation in the ternary fuel, as noted by Kim et. al. While the rate of redistribution has been observed to be slower, numerous metallographs of U-Zr fuel show distinct zone formations.

  10. HTGR Technology Family Assessment for a Range of Fuel Cycle Missions

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Samuel E. Bays; Nick Soelberg

    2010-08-01

    This report examines how the HTGR technology family can provide options for the once through, modified open cycle (MOC), or full recycle fuel cycle strategies. The HTGR can serve all the fuel cycle missions that an LWR can; both are thermal reactors. Additional analyses are warranted to determine if HTGR “full recycle” service could provide improved consumption of transuranic (TRU) material than LWRs (as expected), to analyze the unique proliferation resistance issues associated with the “pebble bed” approach, and to further test and analyze methods to separate TRISO-coated fuel particles from graphite and/or to separate used HTGR fuel meat from its TRISO coating. The feasibility of these two separation issues is not in doubt, but further R&D could clarify and reduce the cost and enable options not adequately explored at present. The analyses here and the now-demonstrated higher fuel burnup tests (after the illustrative designs studied here) should enable future MOC and full recycle HTGR concepts to more rapidly consume TRU, thereby offering waste management advantages. Interest in “limited separation” or “minimum fuel treatment” separation approaches motivates study of impurity-tolerant fuel fabrication. Several issues are outside the scope of this report, including the following: thorium fuel cycles, gas-cooled fast reactors, the reliability of TRISO-coated particles (billions in a reactor), and how soon any new reactor or fuel type could be licensed and then deployed and therefore impact fuel cycle performance measures.

  11. Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carter, W.L.; Olsen, A.R.

    1979-06-01

    The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.

  12. Verification and Validation of the BISON Fuel Performance Code for PCMI Applications

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle Allan Lawrence [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Gardner, Russell James [Idaho National Laboratory; Perez, Danielle Marie [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory

    2016-06-01

    BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described. Validation for application to light water reactor (LWR) PCMI problems is assessed by comparing predicted and measured rod diameter following base irradiation and power ramps. Results indicate a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. Initial rod diameter comparisons have led to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.

  13. Sensitivity analysis on various parameters for lattice analysis of DUPIC fuel with WIMS-AECL code

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok; Park, Jee Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The code WIMS-AECL has been used for the lattice analysis of DUPIC fuel. The lattice parameters calculated by the code is sensitive to the choice of number of parameters, such as the number of tracking lines, number of condensed groups, mesh spacing in the moderator region, other parameters vital to the calculation of probabilities and burnup analysis. We have studied this sensitivity with respect to these parameters and recommend their proper values which are necessary for carrying out the lattice analysis of DUPIC fuel.

  14. Nuclear fuel cycle system simulation tool based on high-fidelity component modeling

    Energy Technology Data Exchange (ETDEWEB)

    Ames, David E.,

    2014-02-01

    The DOE is currently directing extensive research into developing fuel cycle technologies that will enable the safe, secure, economic, and sustainable expansion of nuclear energy. The task is formidable considering the numerous fuel cycle options, the large dynamic systems that each represent, and the necessity to accurately predict their behavior. The path to successfully develop and implement an advanced fuel cycle is highly dependent on the modeling capabilities and simulation tools available for performing useful relevant analysis to assist stakeholders in decision making. Therefore a high-fidelity fuel cycle simulation tool that performs system analysis, including uncertainty quantification and optimization was developed. The resulting simulator also includes the capability to calculate environmental impact measures for individual components and the system. An integrated system method and analysis approach that provides consistent and comprehensive evaluations of advanced fuel cycles was developed. A general approach was utilized allowing for the system to be modified in order to provide analysis for other systems with similar attributes. By utilizing this approach, the framework for simulating many different fuel cycle options is provided. Two example fuel cycle configurations were developed to take advantage of used fuel recycling and transmutation capabilities in waste management scenarios leading to minimized waste inventories.

  15. Extending the world's uranium resources through advanced CANDU fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    De Vuono, Tony; Yee, Frank; Aleyaseen, Val; Kuran, Sermet; Cottrell, Catherine

    2010-09-15

    The growing demand for nuclear power will encourage many countries to undertake initiatives to ensure a self-reliant fuel source supply. Uranium is currently the only fuel utilized in nuclear reactors. There are increasing concerns that primary uranium sources will not be enough to meet future needs. AECL has developed a fuel cycle vision that incorporates other sources of advanced fuels to be adaptable to its CANDU technology.

  16. Application of the BISON Fuel Performance Code to the FUMEX-III Coordinated Research Project

    Energy Technology Data Exchange (ETDEWEB)

    R. L. Williamson; S. R. Novascone

    2012-04-01

    INL recently participated in FUMEX-III, an International Atomic Energy Agency sponsored fuel modeling Coordinated Research Project. A main purpose of FUMEX-III is to compare code predictions to reliable experimental data. During the same time period, the INL initiated development of a new multidimensional (2D and 3D) multiphysics nuclear fuel performance code called BISON. Interactions with international fuel modeling researchers via FUMEX-III played a significant and important role in the BISON evolution, particularly influencing the selection of material and behavioral models which are now included in the code. BISON's ability to model integral fuel rod behavior did not mature until 2011, thus the only FUMEX-III case considered was the Riso3-GE7 experiment, which includes measurements of rod outer diameter following pellet clad mechanical interaction (PCMI) resulting from a power ramp late in fuel life. BISON comparisons to the Riso3-GE7 final rod diameter measurements are quite reasonable. The INL is very interested in participation in the next Fuel Modeling Coordinated Research Project and would like to see the project initiated as soon as possible.

  17. A mono-dimensional nuclear fuel performance analysis code, PUMA, development from a coupled approach

    Energy Technology Data Exchange (ETDEWEB)

    Cheon, J. S.; Lee, B. O.; Lee, C. B. [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong, Daejeon, 305-353 (Korea, Republic of); Yacout, A. M. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2013-07-01

    Multidimensional-multi-physical phenomena in nuclear fuels are treated as a set of mono-dimensional-coupled problems which encompass heat, displacement, fuel constituent redistribution, and fission gas release. Rather than uncoupling these coupled equations as in conventional fuel performance analysis codes, efforts are put into to obtain fully coupled solutions by relying on the recent advances of numerical analysis. Through this approach, a new SFR metal fuel performance analysis code, called PUMA (Performance of Uranium Metal fuel rod Analysis code) is under development. Although coupling between temperature and fuel constituent was made easily, the coupling between the mechanical equilibrium equation and a set of stiff kinetics equations for fission gas release is accomplished by introducing one-level Newton scheme through backward differentiation formula. Displacement equations from 1D finite element formulation of the mechanical equilibrium equation are solved simultaneously with stress equation, creep equation, swelling equation, and FGR equations. Calculations was made successfully such that the swelling and the hydrostatic pressure are interrelated each other. (authors)

  18. Thermodynamic Investigation of an Integrated Gasification Plant with Solid Oxide Fuel Cell and Steam Cycles

    DEFF Research Database (Denmark)

    Rokni, Masoud

    2012-01-01

    A gasification plant is integrated on the top of a solid oxide fuel cell (SOFC) cycle, while a steam turbine (ST) cycle is used as a bottoming cycle for the SOFC plant. The gasification plant was fueled by woodchips to produce biogas and the SOFC stacks were fired with biogas. The produced gas...... was rather clean for feeding to the SOFC stacks after a simple cleaning step. Because all the fuel cannot be burned in the SOFC stacks, a burner was used to combust the remaining fuel. The off-gases from the burner were then used to produce steam for the bottoming steam cycle in a heat recovery steam...... generator (HRSG). The steam cycle was modeled with a simple single pressure level. In addition, a hybrid recuperator was used to recover more energy from the HRSG and send it back to the SOFC cycle. Thus two different configurations were investigated to study the plants characteristic. Such system...

  19. Speciation, in the nuclear fuel cycle by spectroscopic techniques

    Energy Technology Data Exchange (ETDEWEB)

    Colette, S.; Plancque, G.; Allain, F.; Lamouroux, C.; Steiner, V.; Amekraz, B.; Moulin, C. [CEA/Saclay, Dept, des Procedes d' Enrichissement (DPE), 91 - Gif-sur-Yvette (France)

    2000-07-01

    New analytical techniques allowing to perform speciation in the framework of the nuclear fuel cycle are more and more needed. They have to be selective (since matrix encountered are very complex), sensitive (in order to work at representative concentration and below solubility limit), as well as non intrusive (in order to keep the image of the real solution). Among them, laser-based analytical techniques present these advantages together with the possibility to perform remote measurements via fiber optics. Hence, Time-Resolved Laser-Induced Fluorescence (TRLIF) has been used for actinides/lanthanides interaction and speciation studies in inorganic and organic matrices from the reprocessing to waste storage. Moreover, new ion detection methods such as Electro-Spray - Mass Spectrometry (ES-MS) seems promising for speciation studies. Hence, it is the first time that it is possible to directly couple a liquid at atmospheric pressure to a mass detection working at reduced pressure with a soft mode of ionisation that should allow to give informations on chemical species present. Principle, advantages and limitations as well as results obtained with the use of TRLIF and ES-MS on different systems of interest including actinides, lanthanides, fission products in interaction with simple organic molecules to very complex structure will be presented and discussed. (authors)

  20. Managing the Nuclear Fuel Cycle, The Big Picture

    Energy Technology Data Exchange (ETDEWEB)

    Brett W Carlsen

    2010-07-01

    The nuclear industry, at least in the United States, has failed to deliver on its promise of cheap, abundant energy. After pioneering the science and application and becoming a primary exporter of nuclear technologies, domestic use of nuclear power fell out-of-favor with the public and has been relatively stagnant for several decades. Recently, renewed interest has generated optimism and talk of a nuclear renaissance characterized by a new generation of safe, clean nuclear plants in this country. But, as illustrated by recent policy shifts regarding closure of the fuel cycle and geologic disposal of high-level radioactive wastes, significant hurdles have yet to be overcome. Using the principles of system dynamics, this paper will take a holistic look at the nuclear industry and the interactions between the key players to explore both the intended and unintended consequences of efforts to address the issues that have impeded the growth of the industry and also to illustrate aspects which must be effectively addressed if the renaissance of our industry is to be achieved and sustained.

  1. Perform Thermodynamics Measurements on Fuel Cycle Case Study Systems

    Energy Technology Data Exchange (ETDEWEB)

    Martin, Leigh R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    This document was prepared to meet FCR&D level 3 milestone M3FT-14IN0304022, “Perform Thermodynamics Measurements on Fuel Cycle Case Study Systems.” This work was carried out under the auspices of the Thermodynamics and Kinetics FCR&D work package. This document reports preliminary work in support of determining the thermodynamic parameters for the ALSEP process. The ALSEP process is a mixed extractant system comprised of a cation exchanger 2-ethylhexyl-phosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) and a neutral solvating extractant N,N,N’,N’-tetraoctyldiglycolamide (TODGA). The extractant combination produces complex organic phase chemistry that is challenging for traditional measurement techniques. To neutralize the complexity, temperature dependent solvent extraction experiments were conducted with neat TODGA and scaled down concentrations of the ALSEP formulation to determine the enthalpies of extraction for the two conditions. A full set of thermodynamic data for Eu, Am, and Cm extraction by TODGA from 3.0 M HNO3 is reported. These data are compared to previous extraction results from a 1.0 M HNO3 aqueous medium, and a short discussion of the mixed HEH[EHP]/TODGA system results is offered.

  2. Modelling of the Gadolinium Fuel Test IFA-681 using the BISON Code

    Energy Technology Data Exchange (ETDEWEB)

    Pastore, Giovanni [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Williamson, Richard L [Idaho National Laboratory

    2016-05-01

    In this work, application of Idaho National Laboratory’s fuel performance code BISON to modelling of fuel rods from the Halden IFA-681 gadolinium fuel test is presented. First, an overview is given of BISON models, focusing on UO2/UO2-Gd2O3 fuel and Zircaloy cladding. Then, BISON analyses of selected fuel rods from the IFA-681 test are performed. For the first time in a BISON application to integral fuel rod simulations, the analysis is informed by detailed neutronics calculations in order to accurately capture the radial power profile throughout the fuel, which is strongly affected by the complex evolution of absorber Gd isotopes. In particular, radial power profiles calculated at IFE–Halden Reactor Project with the HELIOS code are used. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project. Some slide have been added as an Appendix to present the newly developed PolyPole-1 algorithm for modeling of intra-granular fission gas release.

  3. Modeling and Analysis of FCM UN TRISO Fuel Using the PARFUME Code

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Collin

    2013-09-01

    The PARFUME (PARticle Fuel ModEl) modeling code was used to assess the overall fuel performance of uranium nitride (UN) tri-structural isotropic (TRISO) ceramic fuel in the frame of the design and development of Fully Ceramic Matrix (FCM) fuel. A specific modeling of a TRISO particle with UN kernel was developed with PARFUME, and its behavior was assessed in irradiation conditions typical of a Light Water Reactor (LWR). The calculations were used to access the dimensional changes of the fuel particle layers and kernel, including the formation of an internal gap. The survivability of the UN TRISO particle was estimated depending on the strain behavior of the constituent materials at high fast fluence and burn-up. For nominal cases, internal gas pressure and representative thermal profiles across the kernel and layers were determined along with stress levels in the pyrolytic carbon (PyC) and silicon carbide (SiC) layers. These parameters were then used to evaluate fuel particle failure probabilities. Results of the study show that the survivability of UN TRISO fuel under LWR irradiation conditions might only be guaranteed if the kernel and PyC swelling rates are limited at high fast fluence and burn-up. These material properties are unknown at the irradiation levels expected to be reached by UN TRISO fuel in LWRs. Therefore, more effort is needed to determine them and positively conclude on the applicability of FCM fuel to LWRs.

  4. A study of fuel failure behavior in high burnup HTGR fuel. Analysis by STRESS3 and STAPLE codes

    Energy Technology Data Exchange (ETDEWEB)

    Martin, David G.; Sawa, Kazuhiro; Ueta, Shouhei; Sumita, Junya [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2001-05-01

    In current high temperature gas-cooled reactors (HTGRs), Tri-isotropic coated fuel particles are employed as fuel. In safety design of the HTGR fuels, it is important to retain fission products within particles so that their release to primary coolant does not exceed an acceptable level. From this point of view, the basic design criteria for the fuel are to minimize the failure fraction of as-fabricated fuel coating layers and to prevent significant additional fuel failures during operation. This report attempts to model fuel behavior in irradiation tests using the U.K. codes STRESS3 and STAPLE. Test results in 91F-1A and HRB-22 capsules irradiation tests, which were carried out at the Japan Materials Testing Reactor of JAERI and at the High Flux Isotope Reactor of Oak Ridge National Laboratory, respectively, were employed in the calculation. The maximum burnup and fast neutron fluence were about 10%FIMA and 3 x 10{sup 25} m{sup -2}, respectively. The fuel for the irradiation tests was called high burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the High Temperature Engineering Test Reactor. The calculation results demonstrated that if only mean fracture stress values of PyC and SiC are used in the calculation it is not possible to predict any particle failures, by which is meant when all three load bearing layers have failed. By contrast, when statistical variations in the fracture stresses and particle specifications are taken into account, as is done in the STAPLE code, failures can be predicted. In the HRB-22 irradiation test, it was concluded that the first two particles which had failed were defective in some way, but that the third and fourth failures can be accounted for by the pressure vessel model. In the 91F-1A irradiation test, the result showed that 1 or 2 particles had failed towards the end of irradiation in the upper capsule and no particles failed in the lower capsule. (author)

  5. The cycle of the nuclear fuel used in EDF power plants; Le cycle du combustible nucleaire utilise dans les centrales EDF

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-11-15

    This document briefly indicates the different stages of the nuclear fuel cycle, from the purchase of natural uranium to waste storage. It also indicates the main responsibilities of EDF regarding this fuel cycle (to secure supplies, to organise material transportation, to process and store used fuels and associated wastes). It presents the different associated processes: uranium extraction, purification and concentration, conversion or fluoridation, enrichment. It briefly describes the fuel assembly fabrication, and indicates the main uranium producers in the world. Other addressed steps are: the transportation of fuel assembly, fuel loading, and spent fuel management, the processing of spent fuel and radioactive wastes

  6. Light water reactor fuel analysis code FEMAXI-V (Ver.1)

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-09-01

    A light water fuel analysis code FEMAXI-V is an advanced version which has been produced by integrating FEMAXI-IV(Ver.2), high burn-up fuel code EXBURN-I, and a number of functional improvements and extensions, to predict fuel rod behavior in normal and transient (not accident) conditions. The present report describes in detail the basic theories and structure, models and numerical solutions applied, improvements and extensions, and the material properties adopted in FEMAXI-V(Ver.1). FEMAXI-V deals with a single fuel rod. It predicts thermal and mechanical response of fuel rod to irradiation, including FP gas release. The thermal analysis predicts rod temperature distribution on the basis of pellet heat generation, changes in pellet thermal conductivity and gap thermal conductance, (transient) change in surface heat transfer to coolant, using radial one-dimensional geometry. The heat generation density profile of pellet can be determined by adopting the calculated results of burning analysis code. The mechanical analysis performs elastic/plastic, creep and PCMI calculations by FEM. The FP gas release model calculates diffusion of FP gas atoms and accumulation in bubbles, release and increase in internal pressure of rod. In every analysis, it is possible to allow some materials properties and empirical equations to depend on the local burnup or heat flux, which enables particularly analysis of high burnup fuel behavior and boiling transient of BWR rod. In order to facilitate effective and wide-ranging application of the code, formats and methods of input/output of the code are also described, and a sample output in an actual form is included. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Science.gov (United States)

    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.

  10. Interim assessment of the denatured /sup 233/U fuel cycle: feasibility and nonproliferation characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, L.S.; Bartine, D.E.; Burns, T.J. (eds.)

    1978-12-01

    A fuel cycle that employs /sup 233/U denatured with /sup 238/U and mixed with thorium fertile material is examined with respect to its proliferation-resistance characteristics and its technical and economic feasibility. The rationale for considering the denatured /sup 233/U fuel cycle is presented, and the impact of the denatured fuel on the performance of Light-Water Reactors, Spectral-Shift-Controlled Reactors, Gas-Cooled Reactors, Heavy-Water Reactors, and Fast Breeder Reactors is discussed. The scope of the R, D and D programs to commercialize these reactors and their associated fuel cycles is also summarized and the resource requirements and economics of denatured /sup 233/U cycles are compared to those of the conventional Pu/U cycle. In addition, several nuclear power systems that employ denatured /sup 233/U fuel and are based on the energy center concept are evaluated. Under this concept, dispersed power reactors fueled with denatured or low-enriched uranium fuel are supported by secure energy centers in which sensitive activities of the nuclear cycle are performed. These activities include /sup 233/U production by Pu-fueled transmuters (thermal or fast reactors) and reprocessing. A summary chapter presents the most significant conclusions from the study and recommends areas for future work.

  11. Method for modeling driving cycles, fuel use, and emissions for over snow vehicles.

    Science.gov (United States)

    Hu, Jiangchuan; Frey, H Christopher; Sandhu, Gurdas S; Graver, Brandon M; Bishop, Gary A; Schuchmann, Brent G; Ray, John D

    2014-07-15

    As input to a winter use plan, activity, fuel use, and tailpipe exhaust emissions of over snow vehicles (OSV), including five snow coaches and one snowmobile, were measured on a designated route in Yellowstone National Park (YNP). Engine load was quantified in terms of vehicle specific power (VSP), which is a function of speed, acceleration, and road grade. Compared to highway vehicles, VSP for OSVs is more sensitive to rolling resistance and less sensitive to aerodynamic drag. Fuel use rates increased linearly (R2>0.96) with VSP. For gasoline-fueled OSVs, fuel-based emission rates of carbon monoxide (CO) and nitrogen oxides (NOx) typically increased with increasing fuel use rate, with some cases of very high CO emissions. For the diesel OSVs, which had selective catalytic reduction and diesel particulate filters, fuel-based NOx and particulate matter (PM) emission rates were not sensitive to fuel flow rate, and the emission controls were effective. Inter vehicle variability in cycle average fuel use and emissions rates for CO and NOx was substantial. However, there was relatively little inter-cycle variation in cycle average fuel use and emission rates when comparing driving cycles. Recommendations are made regarding how real-world OSV activity, fuel use, and emissions data can be improved.

  12. Waste Classification based on Waste Form Heat Generation in Advanced Nuclear Fuel Cycles Using the Fuel-Cycle Integration and Tradeoffs (FIT) Model

    Energy Technology Data Exchange (ETDEWEB)

    Denia Djokic; Steven J. Piet; Layne F. Pincock; Nick R. Soelberg

    2013-02-01

    This study explores the impact of wastes generated from potential future fuel cycles and the issues presented by classifying these under current classification criteria, and discusses the possibility of a comprehensive and consistent characteristics-based classification framework based on new waste streams created from advanced fuel cycles. A static mass flow model, Fuel-Cycle Integration and Tradeoffs (FIT), was used to calculate the composition of waste streams resulting from different nuclear fuel cycle choices. This analysis focuses on the impact of waste form heat load on waste classification practices, although classifying by metrics of radiotoxicity, mass, and volume is also possible. The value of separation of heat-generating fission products and actinides in different fuel cycles is discussed. It was shown that the benefits of reducing the short-term fission-product heat load of waste destined for geologic disposal are neglected under the current source-based radioactive waste classification system , and that it is useful to classify waste streams based on how favorable the impact of interim storage is in increasing repository capacity.

  13. User's manual for PRESTO: A computer code for the performance of regenerative steam turbine cycles

    Science.gov (United States)

    Fuller, L. C.; Stovall, T. K.

    1979-01-01

    Standard turbine cycles for baseload power plants and cycles with such additional features as process steam extraction and induction and feedwater heating by external heat sources may be modeled. Peaking and high back pressure cycles are also included. The code's methodology is to use the expansion line efficiencies, exhaust loss, leakages, mechanical losses, and generator losses to calculate the heat rate and generator output. A general description of the code is given as well as the instructions for input data preparation. Appended are two complete example cases.

  14. Analysis of fuel cycle strategies and U.S. transition scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Wigeland, Roald; Taiwo, Temitope A.

    2016-10-17

    The nuclear fuel cycle Evaluation and Screening (E&S) study that was completed in October 2014 [1] enabled the identification of four fuel cycle groups that are considered most promising based on a set of nine evaluation criteria: (a) six benefit criteria of Nuclear Waste Management, Proliferation Risk, Nuclear Material Security Risk, Safety, Environmental Impact, Resource Utilization, and (b) three challenge criteria of Development and Deployment Risk, Institutional Issues, Financial Risk and Economics. The E&S study was conducted at a level of analysis that is "technology- neutral," that is, without consideration of specific technologies, but using the fundamental physics characteristics of each part of the fuel cycle. The study focused on the fuel cycle performance benefits at the fuel cycle equilibrium state, with only limited consideration of transition and deployment impacts. Common characteristics of the four most promising fuel cycle options include continuous recycle of all U/Pu or U/TRU, the use of fast-spectrum reactors, and no use of uranium enrichment once fuel cycle equilibrium has been established. The high-level wastes are mainly from processing of irradiated fuel, and there would be no disposal of any spent fuel. Building on the findings of the E&S study, additional studies have been conducted in the last two years following the information exchange meeting, the 13th IEMPT, which was held in Seoul, the Republic of Korea in 2014. Insights are presented from the recent studies on the benefits and challenges of recycling minor actinides, and transition considerations to some of the most promising fuel cycle options.

  15. Scenarios for the Nuclear fuel cycle in the next decade

    Energy Technology Data Exchange (ETDEWEB)

    Connor, M.J. [Nuclear Resources International, Inc. (NRI), Atlanta, Georgia 30319 (United States); Ortega C, R.F. [Facultad de Ingenieria, UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: NRIAtlanta@aol.com

    2005-07-01

    Some ask: What is the most important event in the Nuclear Fuel Cycle in recent years? One obvious answer is: the dramatic increase in the price of uranium in the international market. The 'spot' or short term delivery price of uranium, increased from $10.90 US dol/lb U{sub 3}O{sub 8} in June 2003 to $14.40 US dol/lb U{sub 3}O{sub 8} in December 2003, a 34% increase in just six months. Then it jumped again to $20.50 US dol/lb U{sub 3}O{sub 8} by December 2004, an increase of 42% that year. Now, by June of 2005, the spot price has climbed another 41% in six months to $29.00 dol/lb U{sub 3}O{sub 8}. Altogether this is a 270% jump in the spot price in just two years. In the same period, the long-term contract price increased from $ 11.75 US dol/lb U{sub 3}O{sub 8} to $30.00 US dol/lb U{sub 3}O{sub 8} - an increase of 255%. These 'adjustments' are a shock to fuel buyers similar to that adjustment of tectonic plates that caused the terrible 'tsunami' in the coast of East Asia last December. This 'adjustment' occurred in a market that most buyers had thought had developed stability - but this was a stability which we now know was mainly due to the supply of large excess inventories of uranium from several countries, including military stocks in the CIS and USA. But what the future holds may be even more dramatic. This paper examines some of the critical elements that will shape the future U{sub 3}O{sub 8} supply/demand relationship, and prices, in the coming decade. (Author)

  16. Fuel assemblies with inert matrices as reloads of cycle 11 of the Unit 1 of the LVNC; Ensamble combustibles con matrices inertes como recargas del ciclo 11 de la Unidad 1 de la CNLV

    Energy Technology Data Exchange (ETDEWEB)

    Lucatero, M.A.; Hernandez M, N.; Hernandez L, H. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mal@nuclear.inin.mx

    2005-07-01

    In this work the results that were obtained of the analysis of three different reloads of the cycle 11 with fuel assemblies containing a mixture of UO{sub 2} and plutonium grade armament in an inert matrix. The proposed assemble, consists of an arrangement 10x10 with 42 bars fuels of PuO{sub 2}-CeO{sub 2}, 34 fuel bars with UO{sub 2} and 16 fuel bars with UO{sub 2}-Gd{sub 2O}3. The proposed assemble is equivalent to an it reloadable assemble of the cycle 11. The fuel bars of uranium and gadolinium, are of the same type of those that are used in the reloadable assemble of uranium. The design and generation of the nuclear databases of the fuel cell with mixed fuel, it was carried out with the HELIUMS code. The simulation of operation of the cycle 11, it was carried out with the CM-PRESTO code. The results show that with one reload of 72 assemblies of UO{sub 2} and 32 assemblies with mixed fuel has a cycle length of smaller in 10.5 days to the cycle length with the complete reload of assemblies of UO{sub 2} and a length smaller cycle in 34 days with the complete reload of 104 assemblies with mixed fuel. (Author)

  17. India's nuclear fuel cycle unraveling the impact of the U.S.-India nuclear accord

    CERN Document Server

    Woddi, Taraknath VK

    2009-01-01

    An analysis of the current (February 2009) status and future potential of India's nuclear fuel cycle is presented in this book. Such a fuel cycle assessment is important, but relatively opaque because India regards various aspects of its nuclear fuel cycle as strategically sensitive. Any study therefore necessarily depends upon reverse calculations based on the information that is available, expert assessments, engineering judgment and anecdotal information. In this work every effort is made to provide transparency to these foundations, so that changes can be made in light of alternative expec

  18. Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

    Energy Technology Data Exchange (ETDEWEB)

    Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong [Kookmin Univ., Seoul (Korea, Republic of)

    2007-03-15

    The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow.

  19. Extension of monodimensional fuel performance codes to finite strain analysis using a Lagrangian logarithmic strain framework

    Energy Technology Data Exchange (ETDEWEB)

    Helfer, Thomas

    2015-07-15

    Highlights: • A simple extension of standard monodimensional fuel performance codes to finite strain is proposed. • Efficiency and reliability are demonstrated. • The logarithmic strain frameword proposed by Miehe et al. is introduced and discussed. - Abstract: This paper shows how the Lagrangian logarithmic strain framework proposed by Miehe et al. can be used to extend monodimensional fuel performance codes, written in the framework of the infinitesimal strain theory, to be able to cope with large deformation of the cladding, such as the ones observed in reactivity initiated accidents (RIA) or loss-of-coolant accidents (LOCA). We demonstrate that the changes only concern the mechanical behaviour integration step by a straightforward modification of the strains (inputs) and the stress (result). The proposed procedure has been implemented in the open-source MFront code generator developed within the PLEIADES platform to handle mechanical behaviours. Using the Alcyone performance code, we apply this procedure to a simulation case proposed within the framework of a recent benchmark on fuel performance codes by the OECD/NEA.

  20. Final Report on Two-Stage Fast Spectrum Fuel Cycle Options

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Won Sik [Purdue Univ., West Lafayette, IN (United States); Lin, C. S. [Purdue Univ., West Lafayette, IN (United States); Hader, J. S. [Purdue Univ., West Lafayette, IN (United States); Park, T. K. [Purdue Univ., West Lafayette, IN (United States); Deng, P. [Purdue Univ., West Lafayette, IN (United States); Yang, G. [Purdue Univ., West Lafayette, IN (United States); Jung, Y. S. [Purdue Univ., West Lafayette, IN (United States); Kim, T. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Stauff, N. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-01-30

    This report presents the performance characteristics of two “two-stage” fast spectrum fuel cycle options proposed to enhance uranium resource utilization and to reduce nuclear waste generation. One is a two-stage fast spectrum fuel cycle option of continuous recycle of plutonium (Pu) in a fast reactor (FR) and subsequent burning of minor actinides (MAs) in an accelerator-driven system (ADS). The first stage is a sodium-cooled FR fuel cycle starting with low-enriched uranium (LEU) fuel; at the equilibrium cycle, the FR is operated using the recovered Pu and natural uranium without supporting LEU. Pu and uranium (U) are co-extracted from the discharged fuel and recycled in the first stage, and the recovered MAs are sent to the second stage. The second stage is a sodium-cooled ADS in which MAs are burned in an inert matrix fuel form. The discharged fuel of ADS is reprocessed, and all the recovered heavy metals (HMs) are recycled into the ADS. The other is a two-stage FR/ADS fuel cycle option with MA targets loaded in the FR. The recovered MAs are not directly sent to ADS, but partially incinerated in the FR in order to reduce the amount of MAs to be sent to the ADS. This is a heterogeneous recycling option of transuranic (TRU) elements

  1. Comparison for thorium fuel cycle facilities of two different capacities for implementation of safeguards

    Energy Technology Data Exchange (ETDEWEB)

    Gangotra, Suresh, E-mail: sgangotra@yahoo.co.in; Grover, R.B.; Ramakumar, K.L.

    2013-09-15

    Highlights: • Facilities for implementation of safeguards for thorium fuel cycle have been compared. • Two concepts have been compared. • In one concept, the facilities are designed in hub and spoke concept. • In second concept the facilities are designed as self-contained concept. • The comparison is done on a number of factors, which affect safeguardability and proliferation resistance. -- Abstract: Thorium based nuclear fuel cycle has many attractive features, its inherent proliferation resistance being one of them. This is due to the presence of high energy gamma emitting daughter products of U{sup 232} associated with U{sup 233}. This high energy gamma radiation also poses challenges in nuclear material accounting. A typical thorium fuel cycle facility has a number of plants including a fuel fabrication plant for initial and equilibrium core, a reprocessed U{sup 233} fuel fabrication plant, a reprocessing plant, a fuel assembly/disassembly plant and associated waste handling and management plants. A thorium fuel cycle facility can be set up to serve reactors at a site. Alternatively, one can follow a hub and spoke approach with a large thorium fuel cycle facility acting as a hub, catering to the requirements of reactors at several sites as spokes. These two concepts have their respective merits and shortcomings in terms of engineering and economics. The present paper is aimed at comparing the merits and challenges for implementation of safeguards on the two concepts viz. a large fuel cycle hub catering to reactors at several sites versus a small fuel cycle facility dedicated to reactors at a single site.

  2. 40 CFR 600.209-08 - Calculation of vehicle-specific 5-cycle fuel economy values for a model type.

    Science.gov (United States)

    2010-07-01

    ...) For alcohol dual fuel automobiles and natural gas dual fuel automobiles, the procedures of paragraphs... from the tests performed using gasoline or diesel test fuel. (ii) If 5-cycle testing was performed on... values for the model type. (5) For alcohol dual fuel automobiles and natural gas dual fuel automobiles...

  3. Studies on supercritical water reactor fuel assemblies using the sub-channel code COBRA-EN

    Energy Technology Data Exchange (ETDEWEB)

    Ammirabile, Luca, E-mail: luca.ammirabile@ec.europa.e [European Commission, JRC, Institute for Energy, Westerduinweg 3, 1755 LE Petten (Netherlands)

    2010-10-15

    In the Generation IV International Forum (GIF) program, the supercritical water reactor (SCWR) concept is among the six innovative reactor types selected for development in the near future. In principle the higher efficiency and better economics make the SCWR concept competitive with the current reactor design. Due to different technical challenges that, however exist, fuel assembly design represents a crucial aspect for the success of this concept. In particular large density variations, low moderation, heat transfer enhancement and deterioration have a strong effect on the core design parameters. Only a few computational tools are currently able to perform sub-channel thermal-hydraulic analysis under supercritical water conditions. At JRC-IE the existing sub-channel code COBRA-EN has been improved to work above the critical pressure of water. The water properties package of the IAPWS Industrial Formulation 1997 was integrated in COBRA-EN to compute the Thermodynamic Properties of Water and Steam. New heat transfer and pressure drop correlations more indicated for the supercritical region of water have also been incorporated in the code. As part of the efforts to appraise the new code capabilities, a code assessment was carried out on the hexagonal fuel assembly of a fast supercritical water reactor. COBRA-EN was also applied in combination with the neutronic code MCNP to investigate on the use of hydride fuel in the HPLWR supercritical water fuel assembly. The results showed that COBRA-EN was able to reproduce the results of similar studies with acceptable accuracy. Future activities will focus on the validation of the code against experimental data and the implementation of new features (counter-current moderator channel, wall, and wire-wrap models).

  4. Fuel Cycle Research and Development Accident Tolerant Fuels Series 1 (ATF-1) Irradiation Testing FY 2016 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Core, Gregory Matthew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    This report contains a summary of irradiation testing of Fuel Cycle Research and Development (FCRD) Accident Tolerant Fuels Series 1 (ATF 1) experiments performed at Idaho National Laboratory (INL) in FY 2016. ATF 1 irradiation testing work performed in FY 2016 included design, analysis, and fabrication of ATF-1B drop in capsule ATF 1 series experiments and irradiation testing of ATF-1 capsules in the ATR.

  5. Social Cost Assessment for Nuclear Fuel Cycle Options in the Republic of Korea

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Ji-eun; Yim, Man-Sung [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    This paper will investigate the vast array of economic factors to estimate the true cost of the nuclear power. There are many studies addressing the external costs of energy production. However, it is only since the 1990s that the external costs of nuclear powered electricity production has been studied in detail. Each investigation has identified their own set of external costs and developed formulas and models using a variety of statistical techniques. The objective of this research is to broaden the scope of the parameters currently consider by adding new areas and expanding on the types of situations considered. Previously the approach to evaluating the external cost of nuclear power did not include various fuel cycle options and influencing parameters. Cost has always been a very important factor in decision-making, in particular for policy choices evaluating the alternative energy sources and electricity generation technologies. Assessment of external costs in support of decision-making should reflect timely consideration of important country specific policy objective. PWR-MOX and FR-Pyro are the best fuel cycle in parameter of environment impacts, but OT or OT-ER is proper than FR-Pyro in human beings. Using the OT fuel cycle is better than FR-Pyro to reduce the conflict cost. When energy supply is deficient, FR-Pyro fuel cycle stands longer than other fuel cycles. Proliferation resistance is shown as 'high' in all fuel cycles, so there are no difference between fuel cycles. When the severe accident occurs, FR-Pyro cycle is economical than other OT based fuel cycles.

  6. Assess How Changes in Fuel Cycle Operation Impact Safeguards

    Energy Technology Data Exchange (ETDEWEB)

    Tobin, Stephen Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Division; Adigun, Babatunde John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Division; Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Division; Trellue, Holly Renee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Division; Sprinkle, James K. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Division

    2016-10-31

    In this report two research topics of interest to Concepts and Approaches are investigated. The motivation of joining them into one project is that both require (1) the simulation of fuel irradiation in a reactor and (2) the transport of gamma and neutron irradiation from the fuel to safeguards detectors. In the next two subsections the merits of each of the two safeguards applications are further introduced. Given the cumulative impact of the enhancements listed above, it is not surprising that both fuel assembly design and fuel assembly irradiation optimization have improved over the past 50+ years. The purpose of the research summarized in this sub-­section is to investigate what, if any, consequence this evolution in reactor operation might have for nuclear safeguards. It is anticipated that the burnup and isotopics of the spent fuel should exhibit less variation over the decades as reactor operators irradiate each assembly to the optimum amount. In contrast, older spent fuel is anticipated to vary more in burnup and resulting isotopics for a given initial enrichment. Continuing with this thesis, modern fuel should be more uniform in composition, and thus, measured safeguards results should be easier to interpret than results from older spent fuel. With spent fuel ponds filling up, interim and long-­term storage of spent fuel will need to be addressed. Additionally after long periods of storage, spent fuel is no longer self-­protecting and as such the IAEA will categorize it as more attractive; in approximately 20 years many of the assemblies from early commercial cores will no longer be considered self-­protecting. This study will assess how more recent changes in the reactor operation could impact the interpretation of safeguards measurements.

  7. Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

    Energy Technology Data Exchange (ETDEWEB)

    E. R. Johnson; R. E. Best

    2009-12-28

    The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the “UREX+3c fuel cycle” and the “Alternative Fuel Cycle” (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount

  8. Comparison of open and closed U-Pu equilibrium fuel cycles for Generation-IV fast reactors with the EQL3D procedure

    Energy Technology Data Exchange (ETDEWEB)

    Krepel, Jiri, E-mail: Jiri.Krepel@psi.ch [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institut, 5232 Villigen PSI (Switzerland); Pelloni, Sandro; Mikityuk, Konstantin [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institut, 5232 Villigen PSI (Switzerland)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer We analyze open and closed fuel cycle of GFR, SFR, and LFR cores by means of ERANOS based EQL3D procedure. Black-Right-Pointing-Pointer Equilibrium of open and closed fuel cycles were compared in terms of their performance and safety parameters. Black-Right-Pointing-Pointer All three cores act in equilibrium closed cycle as iso-breeder and have similar fuel composition. Black-Right-Pointing-Pointer In spite of the same fuel composition the Dopper constants and void reactivities strongly differ between the cores. Black-Right-Pointing-Pointer All three systems seem capable, from neutronics point of view, for the fuel cycle closure. - Abstract: The advanced fast reactors of the fourth generation should enable an indirect burning of poorly fissile {sup 238}U through {sup 239}Pu breeding and recycling of the actinides from their own spent fuel. The recycling or actually the fuel cycle closure can significantly reduce the amount of long-lived radioactive waste and the {sup 238}U burning can multiply the sustainability of the uranium fueled reactors. Regular periodic operation with the fuel recycling converges to an equilibrium cycle. To enable its simulation a numerical tool named equilibrium fuel cycle procedure for fast reactors (EQL3D) was developed in the FAST group of LRS at Paul Scherrer Institut. The procedure is based on the ERANOS code and can be used to yield the description of two basic situations: the equilibrium of an open fuel cycle and the equilibrium of a closed fuel cycle. The goals of the present study are (i) to apply EQL3D to the Gas-cooled Fast Reactor (GFR), Sodium-cooled Fast Reactor (SFR), and Lead-cooled Fast Reactor (LFR), (ii) to simulate and confirm the GFR, SFR, and LFR neutronics capability for closed fuel cycle, and (iii) to evaluate and compare the equilibrium cycle safety and performance parameters. The EQL3D capability enables to characterize the equilibrium cycle for complex reloading patterns

  9. FRAPCON-2: A Computer Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Berna, G. A; Bohn, M. P.; Rausch, W. N.; Williford, R. E.; Lanning, D. D.

    1981-01-01

    FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.

  10. Potential External (non-DOE) Constraints on U.S. Fuel Cycle Options

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet

    2012-07-01

    The DOE Fuel Cycle Technologies (FCT) Program will be conducting a screening of fuel cycle options in FY2013 to help focus fuel cycle R&D activities. As part of this screening, performance criteria and go/no-go criteria are being identified. To help ensure that these criteria are consistent with current policy, an effort was initiated to identify the status and basis of potentially relevant regulations, laws, and policies that have been established external to DOE. As such regulations, laws, and policies may be beyond DOE’s control to change, they may constrain the screening criteria and internally-developed policy. This report contains a historical survey and analysis of publically available domestic documents that could pertain to external constraints on advanced nuclear fuel cycles. “External” is defined as public documents outside DOE. This effort did not include survey and analysis of constraints established internal to DOE.

  11. Study of A Multi-criteria Evaluation Methodology for Nuclear Fuel Cycle System Based on Sustainability

    Institute of Scientific and Technical Information of China (English)

    Liu Jingquan; Hidekazu Yoshikawa; OuYang Jun; Zhou Yangping

    2006-01-01

    This paper presents a multi-criteria evaluation methodology for nuclear fuel cycle options in terms of energy sustainability. Starting from the general sustainability concept and the public acceptance questionnaire, a set of indicators reflecting specific criteria for the evaluation of nuclear fuel cycle options are defined.Particular attention is devoted to the resource utility efficiency, environmental effect, human health hazard and economic effect, which represent the different concerns of different stakeholders. This methodology also integrated a special mathematic processing approach, namely the Extentics Evaluation Method, which quantifies the human being subjective perception to provide the intuitionistic judgement and comparison for different options. The once-through option and reprocessing option of nuclear fuel cycle are examined by using the proposed methodology. The assessment process and result can give us some guidance in nuclear fuel cycle evaluation under the constraint of limited data.

  12. 77 FR 18272 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

    Science.gov (United States)

    2012-03-27

    ... COMMISSION Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...) staff has conducted inspections of the Louisiana Energy Services (LES), LLC, National enrichment... . SUPPLEMENTARY INFORMATION: I. Discussion The NRC staff has conducted inspections of the Louisiana...

  13. Performance of the MTR core with MOX fuel using the MCNP4C2 code.

    Science.gov (United States)

    Shaaban, Ismail; Albarhoum, Mohamad

    2016-08-01

    The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U3O8&PuO2) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U3O8-Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U3O8-Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with (235)U and the amount of loaded (235)U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively.

  14. Preliminary Analysis for Flow Blockage of Plate Fuel using a Commercial CFD Code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong-Pil; Park, Suki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    A plate-type fuel assembly is widely used in the research reactors in order to enhance power density. This type fuel assembly consists of a number of fuel plates, supporting plates, and narrow rectangular cooling channels between the fuel plates. Due to the narrow rectangular channels consisting of parallel plate fuels and side plates, the cooling channels are isolated from each other so that the cross flow between the channels are completely restricted. When blockage at the inlet of the channel occurs, therefore, the coolant flow through the blocked channel will be greatly reduced or completely interrupted. Accordingly, the blocked channel loses its own cooling capability. This event may cause initiation of nucleate boiling and two phase flow instability (FI) in the blocked channel. In recent years, CFD analyses of flow blockage of a research reactor have been performed by many researchers. The present study reports a 3-dimensional CFD simulation using two-phase model for flow blockage of two channels of a plate-type fuel assembly. The present result shows limited information since the simulation is working at the moment. However, the present study indicates that it is possible to predict a damage propagation of a plate-type fuel assembly due to the flow instability caused by flow blockage using a commercial CFD code.

  15. Dynamic Analysis of Nuclear Waste Generation Based on Nuclear Fuel Cycle Transition Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, S. R. [University of Science and Technology, Daejeon (Korea, Republic of); Ko, W. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    According to the recommendations submitted by the Public Engagement Commission on Spent Nuclear Fuel Management (PECOS), the government was advised to pick the site for an underground laboratory and interim storage facilities before the end of 2020 followed by the related research for permanent and underground disposal of spent fuel after 10 years. In the middle of the main issues, the factors of environmentally friendly and safe way to handle nuclear waste are inextricable from nuclear power generating nation to ensure the sustainability of nuclear power. For this purposes, the closed nuclear fuel cycle has been developed regarding deep geological disposal, pyroprocessing, and burner type sodium-cooled fast reactors (SFRs) in Korea. Among two methods of an equilibrium model and a dynamic model generally used for screening nuclear fuel cycle system, the dynamic model is more appropriate to envisage country-specific environment with the transition phase in the long term and significant to estimate meaningful impacts based on the timedependent behavior of harmful wastes. This study aims at analyzing the spent nuclear fuel generation based on the long-term nuclear fuel cycle transition scenarios considered at up-to-date country specific conditions and comparing long term advantages of the developed nuclear fuel cycle option between once-through cycle and Pyro-SFR cycle. In this study, a dynamic analysis was carried out to estimate the long-term projection of nuclear electricity generation, installed capacity, spent nuclear fuel arising in different fuel cycle scenarios based on the up-to-date national energy plans.

  16. International Source Book: Nuclear Fuel Cycle Research and Development Vol 1 Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, K. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lakey, L. T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    1983-07-01

    This document starts with an overview that summarizes nuclear power policies and waste management activities for nations with significant commercial nuclear fuel cycle activities either under way or planned. A more detailed program summary is then included for each country or international agency conducting nuclear fuel cycle and waste management research and development. This first volume includes the overview and the program summaries of those countries listed alphabetically from Argentina to Italy.

  17. Description Fuel Cycle Spanish. Technical Visits; Descripcion del Ciclo de Combustible Espanol. VisitasTecnicas

    Energy Technology Data Exchange (ETDEWEB)

    Ochoa Valero, R.; Vinuesa Carretero, A.

    2014-07-01

    The nuclear fuel cycle includes all processes and operations from the mining of uranium to the management of radioactive waste generated. These processes include the manufacture of nuclear fuel, the operation of the plants and the storage of radioactive waste in the corresponding temporary stores. (Author)

  18. Conception of high safety reactor MAVR, technical and economical fuel cycle characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Kotov, V.M.; Cherepnin, Yu.S.

    1993-12-31

    Operation safety of reactor MAVR under nominal and emergency situations is based on creation of conditions for the minimum time of fuel operation in the core at the minimum quantity of the fissionable material. The variants of core elements construction, of the reactor control systems, and the possible scheme of fuel cycles of the reactor MAVR are considered.

  19. 10 CFR 51.51 - Uranium fuel cycle environmental data-Table S-3.

    Science.gov (United States)

    2010-01-01

    ... with once-through cooling. Fossil fuel: Electrical energy (thousands of MW-hour) 323 Energy 2 2010-01-01 2010-01-01 false Uranium fuel cycle environmental data-Table S-3. 51.51 Section 51.51 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) ENVIRONMENTAL PROTECTION REGULATIONS FOR...

  20. Analysis of fission gas release in LWR fuel using the BISON code

    Energy Technology Data Exchange (ETDEWEB)

    G. Pastore; J.D. Hales; S.R. Novascone; D.M. Perez; B.W. Spencer; R.L. Williamson

    2013-09-01

    Recent advances in the development of the finite-element based, multidimensional fuel performance code BISON of Idaho National Laboratory are presented. Specifically, the development, implementation and testing of a new model for the analysis of fission gas behavior in LWR-UO2 fuel during irradiation are summarized. While retaining a physics-based description of the relevant mechanisms, the model is characterized by a level of complexity suitable for application to engineering-scale nuclear fuel analysis and consistent with the uncertainties pertaining to some parameters. The treatment includes the fundamental features of fission gas behavior, among which are gas diffusion and precipitation in fuel grains, growth and coalescence of gas bubbles at grain faces, grain growth and grain boundary sweeping effects, thermal, athermal, and transient gas release. The BISON code incorporating the new model is applied to the simulation of irradiation experiments from the OECD/NEA International Fuel Performance Experiments database, also included in the IAEA coordinated research projects FUMEX-II and FUMEX-III. The comparison of the results with the available experimental data at moderate burn-up is presented, pointing out an encouraging predictive accuracy, without any fitting applied to the model parameters.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. Fuel burnup analysis for IRIS reactor using MCNPX and WIMS-D5 codes

    Science.gov (United States)

    Amin, E. A.; Bashter, I. I.; Hassan, Nabil M.; Mustafa, S. S.

    2017-02-01

    International Reactor Innovative and Secure (IRIS) reactor is a compact power reactor designed with especial features. It contains Integral Fuel Burnable Absorber (IFBA). The core is heterogeneous both axially and radially. This work provides the full core burn up analysis for IRIS reactor using MCNPX and WIMDS-D5 codes. Criticality calculations, radial and axial power distributions and nuclear peaking factor at the different stages of burnup were studied. Effective multiplication factor values for the core were estimated by coupling MCNPX code with WIMS-D5 code and compared with SAS2H/KENO-V code values at different stages of burnup. The two calculation codes show good agreement and correlation. The values of radial and axial powers for the full core were also compared with published results given by SAS2H/KENO-V code (at the beginning and end of reactor operation). The behavior of both radial and axial power distribution is quiet similar to the other data published by SAS2H/KENO-V code. The peaking factor values estimated in the present work are close to its values calculated by SAS2H/KENO-V code.

  3. POPCYCLE: a computer code for calculating nuclear and fossil plant levelized life-cycle power costs

    Energy Technology Data Exchange (ETDEWEB)

    Hardie, R.W.

    1982-02-01

    POPCYCLE, a computer code designed to calculate levelized life-cycle power costs for nuclear and fossil electrical generating plants is described. Included are (1) derivations of the equations and a discussion of the methodology used by POPCYCLE, (2) a description of the input required by the code, (3) a listing of the input for a sample case, and (4) the output for a sample case.

  4. Life Cycle GHG of NG-Based Fuel and Electric Vehicle in China

    Directory of Open Access Journals (Sweden)

    Qian Zhang

    2013-05-01

    Full Text Available This paper compares the greenhouse gas (GHG emissions of natural gas (NG- based fuels to the GHG emissions of electric vehicles (EVs powered with NG-to-electricity in China. A life-cycle model is used to account for full fuel cycle and use-phase emissions, as well as vehicle cycle and battery manufacturing. The reduction of life-cycle GHG emissions of EVs charged by electricity generated from NG, without utilizing carbon dioxide capture and storage (CCS technology can be 36%–47% when compared to gasoline vehicles. The large range change in emissions reduction potential is driven by the different generation technologies that could in the future be used to generate electricity in China. When CCS is employed in power plants, the GHG emission reductions increase to about 71%–73% compared to gasoline vehicles. It is found that compressed NG (CNG and liquefied NG (LNG fuels can save about 10% of carbon as compared to gasoline vehicles. However, gas-to-liquid (GTL fuel made through the Fischer-Tropsch method will likely lead to a life-cycle GHG emissions increase, potentially 3%–15% higher than gasoline, but roughly equal to petroleum-based diesel. When CCS is utilized, the GTL fueled vehicles emit roughly equal GHG emissions to petroleum-based diesel fuel high-efficient hybrid electric vehicle from the life-cycle perspective.

  5. Model development for quantitative evaluation of proliferation resistance of nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Kim, Ho Dong; Yang, Myung Seung

    2000-07-01

    This study addresses the quantitative evaluation of the proliferation resistance which is important factor of the alternative nuclear fuel cycle system. In this study, model was developed to quantitatively evaluate the proliferation resistance of the nuclear fuel cycles. The proposed models were then applied to Korean environment as a sample study to provide better references for the determination of future nuclear fuel cycle system in Korea. In order to quantify the proliferation resistance of the nuclear fuel cycle, the proliferation resistance index was defined in imitation of an electrical circuit with an electromotive force and various electrical resistance components. The analysis on the proliferation resistance of nuclear fuel cycles has shown that the resistance index as defined herein can be used as an international measure of the relative risk of the nuclear proliferation if the motivation index is appropriately defined. It has also shown that the proposed model can include political issues as well as technical ones relevant to the proliferation resistance, and consider all facilities and activities in a specific nuclear fuel cycle (from mining to disposal). In addition, sensitivity analyses on the sample study indicate that the direct disposal option in a country with high nuclear propensity may give rise to a high risk of the nuclear proliferation than the reprocessing option in a country with low nuclear propensity.

  6. Integrated Fuel-Coolant Interaction (IFCI 7.0) Code User's Manual

    Energy Technology Data Exchange (ETDEWEB)

    Young, Michael F.

    1999-05-01

    The integrated fuel-coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, three-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a description of IFCI 7.0. The user's manual describes the hydrodynamic method and physical models used in IFCI 7.0. Appendix A is an input manual provided for the creation of working decks.

  7. Modeling minor actinide multiple recycling in a lead-cooled fast reactor to demonstrate a fuel cycle without long-lived nuclear waste

    Directory of Open Access Journals (Sweden)

    Stanisz Przemysław

    2015-09-01

    Full Text Available The concept of closed nuclear fuel cycle seems to be the most promising options for the efficient usage of the nuclear energy resources. However, it can be implemented only in fast breeder reactors of the IVth generation, which are characterized by the fast neutron spectrum. The lead-cooled fast reactor (LFR was defined and studied on the level of technical design in order to demonstrate its performance and reliability within the European collaboration on ELSY (European Lead-cooled System and LEADER (Lead-cooled European Advanced Demonstration Reactor projects. It has been demonstrated that LFR meets the requirements of the closed nuclear fuel cycle, where plutonium and minor actinides (MA are recycled for reuse, thereby producing no MA waste. In this study, the most promising option was realized when entire Pu + MA material is fully recycled to produce a new batch of fuel without partitioning. This is the concept of a fuel cycle which asymptotically tends to the adiabatic equilibrium, where the concentrations of plutonium and MA at the beginning of the cycle are restored in the subsequent cycle in the combined process of fuel transmutation and cooling, removal of fission products (FPs, and admixture of depleted uranium. In this way, generation of nuclear waste containing radioactive plutonium and MA can be eliminated. The paper shows methodology applied to the LFR equilibrium fuel cycle assessment, which was developed for the Monte Carlo continuous energy burnup (MCB code, equipped with enhanced modules for material processing and fuel handling. The numerical analysis of the reactor core concerns multiple recycling and recovery of long-lived nuclides and their influence on safety parameters. The paper also presents a general concept of the novel IVth generation breeder reactor with equilibrium fuel and its future role in the management of MA.

  8. A collaboration on extended INPRO case study of the DUPIC fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. H.; Yang, M. S.; Ko, W. I. (and others)

    2007-05-15

    Since 1992, KAERI, AECL, United States Department of States(USDOS) and IAEA have performed the DUPIC fuel cycle development activities as an international cooperative research program, which has now been chosen as a target nuclear system for an INPRO case study. This study will focus on a further improvement and modification of the basic principles, user requirements and acceptance limits, which are defined in the IAEA-TECDOC-1434 for an evaluation of its proliferation-resistance through a proliferation-resistance assessment of the whole fuel cycle of DUPIC based on the INPRO methodology. In order to further develop an evaluation method for a proliferation-resistance based on the INPRO methodology, the basic principles, user requirements and acceptance limits of a proliferation-resistance was reviewed and quantified. Then the evaluation model (material flow, facility scale, reference fuel, etc.) of the DUPIC fuel cycle was developed and a proliferation-resistance assessment of the DUPIC fuel cycle including the PWR fuel cycle was performed by using the revised INPRO methodology in the area of a proliferation resistance. Also, the recommendations for a further improvement of INPRO methodology were suggested through examining the INPRO methodology for a proliferation resistance assessment. Through the proliferation resistance assessment of the whole fuel cycle of DUPIC including the PWR fuel cycle, the proliferation-resistance methodology was updated and re-established. And based on its experience, The research results can be used not only to evaluate and determine the future domestic proliferation-resistant fuel cycles which were derived from the GEN{sub I}V or INPRO programs but also to improve a system design to enhance its proliferation resistance. The present results will be utilized for the development of an INPRO User's Manual which is being developed as an important issue by IAEA. The credibility of the research results were ensured by the IAEA

  9. Feasibility of Burning Civilian Grade Pu in the Modular HTR with Th Fuel Cycle

    Institute of Scientific and Technical Information of China (English)

    许云林; 经荥清

    2001-01-01

    The Modular High Temperature Gas-Cooled Reactor (HTR) can be usedto burn plutonium fuel to reduce Pu stockpiles because of its inherent safety characteristics and ability to burn a variety of fuel mixtures. The equilibrium core is calculated and analyzed for Pu enriched fuel. Fuel spheres with 7 g heavy metal including the civilian grade Pu and thorium are loaded into the reactor. An enrichment of 11% is chosen to provide the desired equilibrium core reactivity. The fuel and moderator temperature coefficients are both negative. The maximum fuel element temperature during normal operation and during a loss of coolant accident is less than 1500 ℃. 92% of 239Pu will be burnt during nomal operation. Therefore, a thorium fuel cycle in the modular HTR is an effective method for burning civilian grade plutonium.

  10. Assessment of PNGV fuels infrastructure. Phase 1 report: Additional capital needs and fuel-cycle energy and emissions impacts

    Energy Technology Data Exchange (ETDEWEB)

    Wang, M.; Stork, K.; Vyas, A.; Mintz, M.; Singh, M.; Johnson, L.

    1997-01-01

    This report presents the methodologies and results of Argonne`s assessment of additional capital needs and the fuel-cycle energy and emissions impacts of using six different fuels in the vehicles with tripled fuel economy (3X vehicles) that the Partnership for a New Generation of Vehicles is currently investigating. The six fuels included in this study are reformulated gasoline, low-sulfur diesel, methanol, ethanol, dimethyl ether, and hydrogen. Reformulated gasoline, methanol, and ethanol are assumed to be burned in spark-ignition, direct-injection engines. Diesel and dimethyl ether are assumed to be burned in compression-ignition, direct-injection engines. Hydrogen and methanol are assumed to be used in fuel-cell vehicles. The authors have analyzed fuels infrastructure impacts under a 3X vehicle low market share scenario and a high market share scenario. The assessment shows that if 3X vehicles are mass-introduced, a considerable amount of capital investment will be needed to build new fuel production plants and to establish distribution infrastructure for methanol, ethanol, dimethyl ether, and hydrogen. Capital needs for production facilities will far exceed those for distribution infrastructure. Among the four fuels, hydrogen will bear the largest capital needs. The fuel efficiency gain by 3X vehicles translates directly into reductions in total energy demand, fossil energy demand, and CO{sub 2} emissions. The combination of fuel substitution and fuel efficiency results in substantial petroleum displacement and large reductions in emissions of nitrogen oxide, carbon monoxide, volatile organic compounds, sulfur oxide, and particulate matter of size smaller than 10 microns.

  11. Development of breached pin performance analysis code SAFFRON (System of Analyzing Failed Fuel under Reactor Operation by Numerical method)

    Energy Technology Data Exchange (ETDEWEB)

    Ukai, Shigeharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1995-03-01

    On the assumption of fuel pin failure, the breached pin performance analysis code SAFFRON was developed to evaluate the fuel pin behavior in relation to the delayed neutron signal response during operational mode beyond the cladding failure. Following characteristic behavior in breached fuel pin is modeled in 3-dimensional finite element method : pellet swelling by fuel-sodium reaction, fuel temperature change, and resultant cladding breach extension and delayed neutron precursors release into coolant. Particularly, practical algorithm of numerical procedure in finite element method was originally developed in order to solve the 3-dimensional non-linear contact problem between the swollen pellet due to fuel-sodium reaction and breached cladding. (author).

  12. Revised Analyses of Decommissioning Reference Non-Fuel-Cycle Facilities

    Energy Technology Data Exchange (ETDEWEB)

    MC Bierschbach; DR Haffner; KJ Schneider; SM Short

    2002-12-01

    Cost information is developed for the conceptual decommissioning of non-fuel-cycle nuclear facilities that represent a significant decommissioning task in terms of decontamination and disposal activities. This study is a re-evaluation of the original study (NUREG/CR-1754 and NUREG/CR-1754, Addendum 1). The reference facilities examined in this study are the same as in the original study and include: a laboratory for the manufacture of {sup 3}H-labeled compounds; a laboratory for the manufacture of {sup 14}C-labeled compounds; a laboratory for the manufacture of {sup 123}I-labeled compounds; a laboratory for the manufacture of {sup 137}Cs sealed sources; a laboratory for the manufacture of {sup 241}Am sealed sources; and an institutional user laboratory. In addition to the laboratories, three reference sites that require some decommissioning effort were also examined. These sites are: (1) a site with a contaminated drain line and hold-up tank; (2) a site with a contaminated ground surface; and (3) a tailings pile containing uranium and thorium residues. Decommissioning of these reference facilities and sites can be accomplished using techniques and equipment that are in common industrial use. Essentially the same technology assumed in the original study is used in this study. For the reference laboratory-type facilities, the study approach is to first evaluate the decommissioning of individual components (e.g., fume hoods, glove boxes, and building surfaces) that are common to many laboratory facilities. The information obtained from analyzing the individual components of each facility are then used to determine the cost, manpower requirements and dose information for the decommissioning of the entire facility. DECON, the objective of the 1988 Rulemaking for materials facilities, is the decommissioning alternative evaluated for the reference laboratories because it results in the release of the facility for restricted or unrestricted use as soon as possible. For a

  13. Land and Water Use, CO2 Emissions, and Worker Radiological Exposure Factors for the Nuclear Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Brett W Carlsen; Brent W Dixon; Urairisa Pathanapirom; Eric Schneider; Bethany L. Smith; Timothy M. AUlt; Allen G. Croff; Steven L. Krahn

    2013-08-01

    The Department of Energy Office of Nuclear Energy’s Fuel Cycle Technologies program is preparing to evaluate several proposed nuclear fuel cycle options to help guide and prioritize Fuel Cycle Technology research and development. Metrics are being developed to assess performance against nine evaluation criteria that will be used to assess relevant impacts resulting from all phases of the fuel cycle. This report focuses on four specific environmental metrics. • land use • water use • CO2 emissions • radiological Dose to workers Impacts associated with the processes in the front-end of the nuclear fuel cycle, mining through enrichment and deconversion of DUF6 are summarized from FCRD-FCO-2012-000124, Revision 1. Impact estimates are developed within this report for the remaining phases of the nuclear fuel cycle. These phases include fuel fabrication, reactor construction and operations, fuel reprocessing, and storage, transport, and disposal of associated used fuel and radioactive wastes. Impact estimates for each of the phases of the nuclear fuel cycle are given as impact factors normalized per unit process throughput or output. These impact factors can then be re-scaled against the appropriate mass flows to provide estimates for a wide range of potential fuel cycles. A companion report, FCRD-FCO-2013-000213, applies the impact factors to estimate and provide a comparative evaluation of 40 fuel cycles under consideration relative to these four environmental metrics.

  14. Development of core fuel management code system for WWER-type reactors

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    In this article, a core fuel management program for hexagonal pressurized water type WWER reactors (CFMHEX) has been developed, which is based on advanced three-dimensional nodal method and integrated with thermal hydraulic code to realize the coupling of neutronics and thermal-hydraulics. In CFMHEX, all these feedback effects such as burnup, power distribution, moderator density, and control rod insertion are considered. The verification and validation of the code system have been examined through the IAEA WWER-1000-type Kalinin NPP benchmark problem. The numerical results are in good agreement with measurements and are close to those of other international institutes.

  15. Research on Fuel Consumption of Hybrid Bulldozer under Typical Duty Cycle

    Science.gov (United States)

    Song, Qiang; Wang, Wen-Jun; Jia, Chao; Yao, You-Liang; Wang, Sheng-Bo

    The hybrid drive bulldozer adopts a dual-motor independent drive system with engine-generator assembly as its power source. The mathematical model of the whole system is constructed on the software platform of MATLAB/Simulink. And then according to the velocity data gained from a real test experiment, a typical duty cycle is build up. Finally the fuel consumption of the bulldozer is calculated under this duty-cycle. Simulation results show that, compared with the traditional mechanical one, the hybrid electric drive system can save fuel up to 16% and therefore indicates great potential for lifting up fuel economy.

  16. A Characteristics-Based Approach to Radioactive Waste Classification in Advanced Nuclear Fuel Cycles

    Science.gov (United States)

    Djokic, Denia

    The radioactive waste classification system currently used in the United States primarily relies on a source-based framework. This has lead to numerous issues, such as wastes that are not categorized by their intrinsic risk, or wastes that do not fall under a category within the framework and therefore are without a legal imperative for responsible management. Furthermore, in the possible case that advanced fuel cycles were to be deployed in the United States, the shortcomings of the source-based classification system would be exacerbated: advanced fuel cycles implement processes such as the separation of used nuclear fuel, which introduce new waste streams of varying characteristics. To be able to manage and dispose of these potential new wastes properly, development of a classification system that would assign appropriate level of management to each type of waste based on its physical properties is imperative. This dissertation explores how characteristics from wastes generated from potential future nuclear fuel cycles could be coupled with a characteristics-based classification framework. A static mass flow model developed under the Department of Energy's Fuel Cycle Research & Development program, called the Fuel-cycle Integration and Tradeoffs (FIT) model, was used to calculate the composition of waste streams resulting from different nuclear fuel cycle choices: two modified open fuel cycle cases (recycle in MOX reactor) and two different continuous-recycle fast reactor recycle cases (oxide and metal fuel fast reactors). This analysis focuses on the impact of waste heat load on waste classification practices, although future work could involve coupling waste heat load with metrics of radiotoxicity and longevity. The value of separation of heat-generating fission products and actinides in different fuel cycles and how it could inform long- and short-term disposal management is discussed. It is shown that the benefits of reducing the short-term fission

  17. Detailed description and user`s manual of high burnup fuel analysis code EXBURN-I

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Saitou, Hiroaki

    1997-11-01

    EXBURN-I has been developed for the analysis of LWR high burnup fuel behavior in normal operation and power transient conditions. In the high burnup region, phenomena occur which are different in quality from those expected for the extension of behaviors in the mid-burnup region. To analyze these phenomena, EXBURN-I has been formed by the incorporation of such new models as pellet thermal conductivity change, burnup-dependent FP gas release rate, and cladding oxide layer growth to the basic structure of low- and mid-burnup fuel analysis code FEMAXI-IV. The present report describes in detail the whole structure of the code, models, and materials properties. Also, it includes a detailed input manual and sample output, etc. (author). 55 refs.

  18. Evaluation of CASMO-3 and HELIOS for Fuel Assembly Analysis from Monte Carlo Code

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Hyung Jin; Song, Jae Seung; Lee, Chung Chan

    2007-05-15

    This report presents a study comparing deterministic lattice physics calculations with Monte Carlo calculations for LWR fuel pin and assembly problems. The study has focused on comparing results from the lattice physics code CASMO-3 and HELIOS against those from the continuous-energy Monte Carlo code McCARD. The comparisons include k{sub inf}, isotopic number densities, and pin power distributions. The CASMO-3 and HELIOS calculations for the k{sub inf}'s of the LWR fuel pin problems show good agreement with McCARD within 956pcm and 658pcm, respectively. For the assembly problems with Gadolinia burnable poison rods, the largest difference between the k{sub inf}'s is 1463pcm with CASMO-3 and 1141pcm with HELIOS. RMS errors for the pin power distributions of CASMO-3 and HELIOS are within 1.3% and 1.5%, respectively.

  19. A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

    Science.gov (United States)

    Karahan, Aydın; Buongiorno, Jacopo

    2010-01-01

    An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors

  20. A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydin, E-mail: karahan@mit.ed [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology (United States); Buongiorno, Jacopo [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology (United States)

    2010-01-31

    An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO{sub 2}-PuO{sub 2} mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium

  1. Calculational Benchmark Problems for VVER-1000 Mixed Oxide Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Emmett, M.B.

    2000-03-17

    Standard problems were created to test the ability of American and Russian computational methods and data regarding the analysis of the storage and handling of Russian pressurized water reactor (VVER) mixed oxide fuel. Criticality safety and radiation shielding problems were analyzed. Analysis of American and Russian multiplication factors for fresh fuel storage for low-enriched uranium (UOX), weapons- (MOX-W) and reactor-grade (MOX-R) MOX differ by less than 2% for all variations of water density. For shielding calculations for fresh fuel, the ORNL results for the neutron source differ from the Russian results by less than 1% for UOX and MOX-R and by approximately 3% for MOX-W. For shielding calculations for fresh fuel assemblies, neutron dose rates at the surface of the assemblies differ from the Russian results by 5% to 9%; the level of agreement for gamma dose varies depending on the type of fuel, with UOX differing by the largest amount. The use of different gamma group structures and instantaneous versus asymptotic decay assumptions also complicate the comparison. For the calculation of dose rates from spent fuel in a shipping cask, the neutron source for UOX after 3-year cooling is within 1% and for MOX-W within 5% of one of the Russian results while the MOX-R difference is the largest at over 10%. These studies are a portion of the documentation required by the Russian nuclear regulatory authority, GAN, in order to certify Russian programs and data as being acceptably accurate for the analysis of mixed oxide fuels.

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

    Directory of Open Access Journals (Sweden)

    Hangbok Choi

    2013-01-01

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

  3. The life cycle assessment of alternative fuel chains for urban buses and trolleybuses.

    Science.gov (United States)

    Kliucininkas, L; Matulevicius, J; Martuzevicius, D

    2012-05-30

    This paper describes a comparative analysis of public transport alternatives in the city of Kaunas, Lithuania. An LCA (Life Cycle Assessment) inventory analysis of fuel chains was undertaken using the midi urban bus and a similar type of trolleybus. The inventory analysis of fuel chains followed the guidelines provided by the ISO 14040 and ISO 14044 standards. The ReCiPe Life Cycle Impact Assessment (LCIA) methodology was used to quantify weighted damage originating from five alternative fuel chains. The compressed biogas fuel chain had the lowest weighted damage value, namely 45.7 mPt/km, whereas weighted damage values of the fuel chains based on electricity generation for trolleybuses were 60.6 mPt/km (for natural gas) and 78.9 mPt/km (for heavy fuel oil). The diesel and compressed natural gas fuel chains exhibited considerably higher damage values of 114.2 mPt/km and 132.6 mPt/km, respectively. The comparative life cycle assessment of fuel chains suggested that biogas-powered buses and electric trolleybuses can be considered as the best alternatives to use when modernizing the public transport fleet in Kaunas.

  4. Optimization of fuel cycles: marginal loss values; Optimisation des cycles de combustibles: valeurs marginales des pertes

    Energy Technology Data Exchange (ETDEWEB)

    Gaussens, J. [Commissariat a l' Energie Atomique, 75 - Paris (France); Lasteyrie, B. de; Doumerc, J. [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques, 75 - Paris (France)

    1965-07-01

    Uranium processing from the pit to the fuel element rod entails metal losses at every step. These losses become more and more expensive with the elaboration of the metal. Some of the uranium must be accepted as definitely lost whilst the rest could be recovered and recycled. The high cost of these losses, whether they are recycled or not, and the fact that the higher the enrichment is the higher their costs are, make it necessary to take them into account when optimizing fuel cycles. It is therefore felt important to determine their most desirable level from an economic point of view at the various nuclear fuel processing stages. However, in France as in some other countries, fissile material production is a state concern, whilst fuel element fabrication is carried out by private enterprise. Optimization criteria and the economic value of losses are therefore different for each of the two links in the fabrication chain. One can try in spite of this to reach an optimum which would conform to public interest, without interfering with the firm's sales policy. This entails using the fact that for a given output marginal costs are equal at the optimum. One can therefore adjust the level of the losses to attain this equation of marginal costs, as these are easier to obtain from the firm than a justification of the actual prices. One notices moreover that, although mainly concerned with losses, this global analysis can bring both the state and the firm to a better use of other production factors. An account is given of the theory of this economic optimization method and practical applications in the field of natural uranium-graphite moderated and CO{sub 2} cooled reactor fuel element fabrication are offered. (authors) [French] L'elaboration de l'uranium, de la mine a la cartouche d'element combustible, s'accompagne, a tous les stades, de pertes, de plus en plus onereuses, a mesure que le metal devient plus elabore. Une certaine proportion de l

  5. LIFE CYCLE BASED STUDIES ON BIOETHANOL FUEL FOR SUSTAINABLE TRANSPORTATION: A LITERATURE REVIEW

    Science.gov (United States)

    A literature search was conducted and revealed 45 publications (1996-2005) that compare bio-ethanol systems to conventional fuel on a life-cycle basis, or using life cycle assessment. Feedstocks, such as sugar beets, wheat, potato, sugar cane, and corn, have been investigated in...

  6. Measurements of Fission Cross Sections for the Isotopes relevant to the Thorium Fuel Cycle

    CERN Multimedia

    2002-01-01

    The present concern about a sustainable energy supply is characterised by a considerable uncertainty: the green house effect and foreseeable limits in fossil fuel resources on the one hand, the concern about the environmental impact of nuclear fission energy and the long term fusion research on the other hand, have led to the consideration of a variety of advanced strategies for the nuclear fuel cycle and related nuclear energy systems. The present research directories concern such strategies as the extension of the life span of presently operating reactors, the increase of the fuel burn-up, the plutonium recycling, and in particular the incineration of actinides and long-Lived fission products, the accelerator driven systems (ADS), like the "Energy Amplifier" (EA) concept of C. Rubbia, and the possible use of the Thorium fuel cycle. The detailed feasibility study and safety assessment of these strategies requires the accurate knowledge of neutron nuclear reaction data. Both, higher fuel burn-up and especiall...

  7. Improved analysis on multiple recycling of fuel in prototype fast breeder reactor in a closed fuel cycle

    Indian Academy of Sciences (India)

    G Pandikumar; V Gopalakrishnan; P Mohanakrishnan

    2011-08-01

    An FBR closed fuel cycle involves recycling of the discharge fuel, after reprocessing and refabrication, to utilize the unburnt fuel remains and the freshly bred fissile material. Our previous study in this regard for the PFBR indicated a comfortable feasibility of multiple recycling with selfsufficiency. In the present work, more refined estimations are done using the most recent nuclear data, viz. ENDF/B-VII.0, and with the most recent specification of the fuel composition. Among others, this paper brings out the importance of taking into account the energy self-shielding effects in the cross-section averages used in the study. While self-shielded averages lead to realistic predictions, unshielded averages significantly overpredict breeding in the blankets and underpredict loss in the cores.

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

    Energy Technology Data Exchange (ETDEWEB)

    Shamasundar, B.I.; Fehrenbach, M.E.

    1981-05-01

    The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations.

  9. Development and verification of NRC`s single-rod fuel performance codes FRAPCON-3 AND FRAPTRAN

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, C.E.; Cunningham, M.E.; Lanning, D.D. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    The FRAPCON and FRAP-T code series, developed in the 1970s and early 1980s, are used by the US Nuclear Regulatory Commission (NRC) to predict fuel performance during steady-state and transient power conditions, respectively. Both code series are now being updated by Pacific Northwest National Laboratory to improve their predictive capabilities at high burnup levels. The newest versions of the codes are called FRAPCON-3 and FRAPTRAN. The updates to fuel property and behavior models are focusing on providing best estimate predictions under steady-state and fast transient power conditions up to extended fuel burnups (> 55 GWd/MTU). Both codes will be assessed against a data base independent of the data base used for code benchmarking and an estimate of code predictive uncertainties will be made based on comparisons to the benchmark and independent data bases.

  10. Analysis of the second part of the fuel cycle of nuclear spanish park using module TR{sub E}VOL of EVOLCODE2; Simulacion del ciclo de combustible del parque nuclear espanol con el codigo TR{sub E}VOL.

    Energy Technology Data Exchange (ETDEWEB)

    Merino Rodriguez, I.; Alvarez-Velarde, F.; Martin-Fuertes, F.

    2011-07-01

    This paper describes the application of the code TR{sub E}VOL an associated fuel cycle Spanish nuclear park, with the objective of estimating the mass of nuclear fuel manufactured by reactor and the mass generated of irradiated fuel.

  11. Synthetic fuel production via carbon neutral cycles with high temperature nuclear reactors as a power source

    Energy Technology Data Exchange (ETDEWEB)

    Konarek, E.; Coulas, B.; Sarvinis, J. [Hatch Ltd., Mississauga, Ontario (Canada)

    2016-06-15

    This paper analyzes a number of carbon neutral cycles, which could be used to produce synthetic hydrocarbon fuels. Synthetic hydrocarbons are produced via the synthesis of Carbon Monoxide and Hydrogen. The . cycles considered will either utilize Gasification processes, or carbon capture as a source of feed material. In addition the cycles will be coupled to a small modular Nuclear Reactor (SMR) as a power and heat source. The goal of this analysis is to reduce or eliminate the need to transport diesel and other fossil fuels to remote regions and to provide a carbon neutral, locally produced hydrocarbon fuel for remote communities. The technical advantages as well as the economic case are discussed for each of the cycles presented. (author)

  12. The long-term carbon cycle, fossil fuels and atmospheric composition.

    Science.gov (United States)

    Berner, Robert A

    2003-11-20

    The long-term carbon cycle operates over millions of years and involves the exchange of carbon between rocks and the Earth's surface. There are many complex feedback pathways between carbon burial, nutrient cycling, atmospheric carbon dioxide and oxygen, and climate. New calculations of carbon fluxes during the Phanerozoic eon (the past 550 million years) illustrate how the long-term carbon cycle has affected the burial of organic matter and fossil-fuel formation, as well as the evolution of atmospheric composition.

  13. Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

    Energy Technology Data Exchange (ETDEWEB)

    E. R. Johnson; R. E. Best

    2009-12-28

    The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the “UREX+3c fuel cycle” and the “Alternative Fuel Cycle” (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount

  14. KALIMER-600-clad Core Fuel Assembly Calculation using MATRA-LMR (V2.0) Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Gyun; Kim, Young Il

    2006-12-15

    Since the sodium boiling point is very high, maximum cladding and pin temperatures are used for design limit condition in sodium cooled liquid metal reactor. It is necessary to predict accurately the temperature distribution in the core and in the subassemblies to increase the sodium coolant efficiency. Based on the MATRA code, which is developed for PWR analysis, MATRA-LMR has been developed for SFR. The major modifications are: the sodium properties table is implemented as subprogram in the code, Heat transfer coefficients are changed for SFR, te pressure drop correlations are changed for more accurate calculations, which are Novendstern, Chiu-Rohsenow-Todreas, and Cheng-Todreas correlations. This This report describes briefly code structure and equations of MATRA-LMR (Version 2.0), explains input data preparation and shows some calculation results for the KALIMER-600-clad core fuel assembly for which has been performed the conceptual design of the core in the year 2006.

  15. Hybrid life-cycle assessment of natural gas based fuel chains for transportation.

    Science.gov (United States)

    Strømman, Anders Hammer; Solli, Christian; Hertwich, Edgar G

    2006-04-15

    This research compares the use of natural gas, methanol, and hydrogen as transportation fuels. These three fuel chains start with the extraction and processing of natural gas in the Norwegian North Sea and end with final use in Central Europe. The end use is passenger transportation with a sub-compact car that has an internal combustion engine for the natural gas case and a fuel cell for the methanol and hydrogen cases. The life cycle assessment is performed by combining a process based life-cycle inventory with economic input-output data. The analysis shows that the potential climate impacts are lowest for the hydrogen fuel scenario with CO2 deposition. The hydrogen fuel chain scenario has no significant environmental disadvantage compared to the other fuel chains. Detailed analysis shows that the construction of the car contributes significantly to most impact categories. Finally, it is shown how the application of a hybrid inventory model ensures a more complete inventory description compared to standard process-based life-cycle assessment. This is particularly significant for car construction which would have been significantly underestimated in this study using standard process life-cycle assessment alone.

  16. Outline of results of safety research (in nuclear fuel cycle field in fiscal year 1996)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    The safety research in Power Reactor and Nuclear Fuel Development Corporation in fiscal year 1996 has been carried out based on the basic plan of safety research (from fiscal year 1996 to 2000) which was decided in March, 1996. In this report, on nuclear fuel cycle field, namely all the subjects in the fields of nuclear fuel facilities, environmental radioactivity and waste disposal, and the subjects related to nuclear fuel facilities among the fields of aseismatic and probabilistic safety assessments, the results of research in fiscal year 1996, the first year of the 5-year project, are summarized together with the outline of the basic plan of safety research. The basic policy, objective and system for promotion of the safety research are described. The objectives of the safety research are the advancement of safety technology, the safety of facilities, stable operation techniques, the safety design and the evaluation techniques of next generation facilities, and the support of transferring nuclear fuel cycle to private businesses. The objects of the research are uranium enrichment, fuel fabrication and reprocessing, and waste treatment and storage. 52 investigation papers of the results of the safety research in nuclear fuel cycle field in fiscal year 1996 are collected in this report. (K.I.)

  17. Operation of CANDU power reactor in thorium self-sufficient fuel cycle

    Indian Academy of Sciences (India)

    B R Bergelson; A S Gerasimov; G V Tikhomirov

    2007-02-01

    This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the same as for standard operation in uranium cycle. Two modes of operations are discussed in the paper: mode of preliminary accumulation of 233U and mode of operation in self-sufficient cycle. For the mode of accumulation of 233U it was assumed for calculations that plutonium can be used as additional fissile material to provide neutrons for 233U production. Plutonium was placed in fuel channels, while 232Th was located in target channels. Maximum content of 233U in target channels was estimated to be ∼ 13 kg/t of ThO2. This was achieved by irradiation for six years. The start of the reactor operation in the self-sufficient mode requires 233U content to be not less than 12 kg/t. For the mode of operation in self-sufficient cycle, it was assumed that all channels were loaded with identical fuel assemblies containing ThO2 and certain amount of 233U. It is shown that nonuniform distribution of 233U in fuel assembly is preferable.

  18. Analysis of factors affecting the implementation of back-end nuclear fuel cycle policy in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yung Myung; Yang, Maeng Ho; Kim, Hyun Joon; Chung, Hwan Sam; Oh, Keun Bae; Lee, Byung OoK; Ko, Han Suk; Song, Ki Dong; Lee, Man Ki; Moon, Ki Hwan; Lee, Han Myung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-01-01

    In this study, the back-end nuclear fuel cycle acceptability is surveyed and analyzed in the following three aspects. To begin with, the future political situation and energy-environmental issues are analyzed as part of the socio-economic aspect. Secondly, the domestic situation of nuclear industries and the fuel cycle policy of foreign countries are surveyed as the technical aspect. Finally, NPT, IAEA safeguards and nuclear export control regimes are analyzed as the institutional aspect. The unification period of South and North Korea also will greatly affect the implementation of back-end fuel cycle policy, and public attitudes will affect the acquisition of site, construction, and operation of nuclear facilities. An effort to release international restrictions on the back-end fuel cycle is also required to accelerate the implementation of the policy. In this regard, the back-end fuel cycle policy should be clear-cut to avoid misunderstanding with respect to nuclear proliferation. Importantly, agreements with foreign countries should be amended at a mutual equivalent level. (Author) 30 refs., 5 figs., 25 tabs.

  19. The mode of operation of CANDU power reactor in thorium self-sufficient fuel cycle

    Directory of Open Access Journals (Sweden)

    Bergelson Boris R.

    2008-01-01

    Full Text Available This paper presents the results of calculations for CANDU reactor operation in the thorium fuel cycle. The calculations were performed to estimate feasibility of operation of a heavy-water thermal neutron power reactor in the self-sufficient thorium cycle. The parameters of the active core and the scheme of fuel reloading were considered to be the same as for the standard operation in the uranium cycle. Two modes of operation are discussed in the paper: the mode of preliminary accumulation of 233U and the mode of operation in the self-sufficient cycle. For calculations for the mode of accumulation of 233U, it was assumed that plutonium was used as the additional fissile material to provide neutrons for 233U production. Plutonium was placed in fuel channels, while 232Th was located in target channels. The maximum content of 233U in the target channels was about 13 kg/t of ThO2. This was achieved by six year irradiation. The start of reactor operation in the self-sufficient mode requires content of 233U not less than 12 kg/t. For the mode of operation in the self-sufficient cycle, it was assumed that all the channels were loaded with the identical fuel assemblies containing ThO2 and a certain amount of 233U. It was shown that the non-uniform distribution of 233U in a fuel assembly is preferable.

  20. Life cycle assessment of the use of alternative fuels in cement kilns: A case study.

    Science.gov (United States)

    Georgiopoulou, Martha; Lyberatos, Gerasimos

    2017-07-14

    The benefits of using alternative fuels (AFs) in the cement industry include reduction of the use of non-renewable fossil fuels and lower emissions of greenhouse gases, since fossil fuels are replaced with materials that would otherwise be degraded or incinerated with corresponding emissions and final residues. Furthermore, the use of alternative fuels maximizes the recovery of energy. Seven different scenaria were developed for the production of 1 ton of clinker in a rotary cement kiln. Each of these scenaria includes the use of alternative fuels such as RDF (Refuse derived fuel), TDF (Tire derived fuel) and BS (Biological sludge) or a mixture of them, in partial replacement of conventional fuels such as coal and pet coke. The purpose of this study is to evaluate the environmental impacts of the use of alternative fuels in relation to conventional fuels in the kiln operation. The Life Cycle Assessment (LCA) methodology is used to quantify the potential environmental impacts in each scenario. The interpretation of the results provides the conclusion that the most environmentally friendly prospect is the scenario based on RDF while the less preferable scenario is the scenario based on BS. Copyright © 2017 Elsevier Ltd. All rights reserved.

  1. Life cycle assessment of biomass-to-liquid fuels - Final report

    Energy Technology Data Exchange (ETDEWEB)

    Jungbluth, N.; Buesser, S.; Frischknecht, R.; Tuchschmid, M.

    2008-02-15

    This study elaborates a life cycle assessment of using of BTL-fuels (biomass-to-liquid). This type of fuel is produced in synthesis process from e.g. wood, straw or other biomass. The life cycle inventory data of the fuel provision with different types of conversion concepts are based on the detailed life cycle assessment compiled and published within a European research project. The inventory of the fuel use emissions is based on information published by automobile manufacturers on reductions due to the use of BTL-fuels. Passenger cars fulfilling the EURO3 emission standards are the basis for the comparison. The life cycle inventories of the use of BTL-fuels for driving in passenger cars are investigated from cradle to grave. The full life cycle is investigated with the transportation of one person over one kilometre (pkm) as a functional unit. This includes all stages of the life cycle of a fuel (biomass and fuel production, distribution, combustion) and the necessary infrastructure (e.g. tractors, conversion plant, cars and streets). The use of biofuels is mainly promoted for the reason of reducing the climate change impact and the use of scarce non-renewable resources e.g. crude oil. The possible implementation of BTL-fuel production processes would potentially help to achieve this goal. The emissions of greenhouse gases due to transport services could be reduced by 28% to 69% with the BTL-processes using straw, forest wood or short-rotation wood as a biomass input. The reduction potential concerning non-renewable energy resources varies between 37% und 61%. A previous study showed that many biofuels cause higher environmental impacts than fossil fuels if several types of ecological problems are considered. The study uses two single score impact assessment methods for the evaluation of the overall environmental impacts, namely the Eco-indicator 99 (H,A) and the Swiss ecological scarcity 2006 method. The transportation with the best BTL-fuel from short

  2. Extending the application range of a fuel performance code from normal operating to design basis accident conditions

    Science.gov (United States)

    Van Uffelen, P.; Győri, C.; Schubert, A.; van de Laar, J.; Hózer, Z.; Spykman, G.

    2008-12-01

    Two types of fuel performance codes are generally being applied, corresponding to the normal operating conditions and the design basis accident conditions, respectively. In order to simplify the code management and the interface between the codes, and to take advantage of the hardware progress it is favourable to generate a code that can cope with both conditions. In the first part of the present paper, we discuss the needs for creating such a code. The second part of the paper describes an example of model developments carried out by various members of the TRANSURANUS user group for coping with a loss of coolant accident (LOCA). In the third part, the validation of the extended fuel performance code is presented for LOCA conditions, whereas the last section summarises the present status and indicates needs for further developments to enable the code to deal with reactivity initiated accident (RIA) events.

  3. Hydrocarbon bio-jet fuel from bioconversion of poplar biomass: life cycle assessment.

    Science.gov (United States)

    Budsberg, Erik; Crawford, Jordan T; Morgan, Hannah; Chin, Wei Shan; Bura, Renata; Gustafson, Rick

    2016-01-01

    Bio-jet fuels compatible with current aviation infrastructure are needed as an alternative to petroleum-based jet fuel to lower greenhouse gas emissions and reduce dependence on fossil fuels. Cradle to grave life cycle analysis is used to investigate the global warming potential and fossil fuel use of converting poplar biomass to drop-in bio-jet fuel via a novel bioconversion platform. Unique to the biorefinery designs in this research is an acetogen fermentation step. Following dilute acid pretreatment and enzymatic hydrolysis, poplar biomass is fermented to acetic acid and then distilled, hydroprocessed, and oligomerized to jet fuel. Natural gas steam reforming and lignin gasification are proposed to meet hydrogen demands at the biorefineries. Separate well to wake simulations are performed using the hydrogen production processes to obtain life cycle data. Both biorefinery designs are assessed using natural gas and hog fuel to meet excess heat demands. Global warming potential of the natural gas steam reforming and lignin gasification bio-jet fuel scenarios range from CO2 equivalences of 60 to 66 and 32 to 73 g MJ(-1), respectively. Fossil fuel usage of the natural gas steam reforming and lignin gasification bio-jet fuel scenarios range from 0.78 to 0.84 and 0.71 to 1.0 MJ MJ(-1), respectively. Lower values for each impact category result from using hog fuel to meet excess heat/steam demands. Higher values result from using natural gas to meet the excess heat demands. Bio-jet fuels produced from the bioconversion of poplar biomass reduce the global warming potential and fossil fuel use compared with petroleum-based jet fuel. Production of hydrogen is identified as a major source of greenhouse gas emissions and fossil fuel use in both the natural gas steam reforming and lignin gasification bio-jet simulations. Using hog fuel instead of natural gas to meet heat demands can help lower the global warming potential and fossil fuel use at the biorefineries.

  4. Numerical design of the Seed-Blanket Unit for the thorium nuclear fuel cycle

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2016-01-01

    Full Text Available In the paper we present the Monte Carlo modelling by the means of the Monte Carlo Continuous Energy Burn-up Code of the 17x17 Pressurized Water Reactor fuel assembly designed according to the Radkowsky Thorium Fuel concept. The design incorporates the UO2 seed fuel located in the centre and (Th,UO2 blanket fuel located in the peripheries of fuel assembly. The high power seed region supplies neutrons for the low power blanket region and thus induces breeding of fissile 233U from fertile 232Th. The both regions are physically separated and thus this approach is also known as either the heterogonous approach or the Seed-Blanket Unit. In the numerical analysis we consider the time evolutions of infinite neutron multiplication factor, axial/radial power density profile, 233U, 235U and 232Th.

  5. Molten Salt Fuel Cycle Requirements for ADTT Applications

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D. F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Del Cul, G. D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Toth, L. M. [Commodore Advanced Sciences, Inc., Oak Ridge, TN (United States)

    1999-06-07

    The operation of an ADT system with the associated nuclear reactions has a profound effect upon the chemistry of the fuel - especially with regards to container compatibility and the chemical separations that may be required. The container can be protected by maintaining the redox chemistry within a relatively narrow, non-corrosive window. Neutron economy as well as other factors require a sophisticated regime of fission product separations. Neither of these control requirements has been demonstrated on the scale or degree of sophistication necessary to support an ADT device. We review the present situation with respect to fluoride salts, and focus on the critical issues in these areas which must be addressed. One requirement for advancement in this area - a supply of suitable materials - will soon be fulfilled by the remediation of ORNL's Molten Salt Reactor Experiment, and the removal of a total of 11,000 kg of enriched (Li-7 > 99.9%) coolant, flush, and fuel salts.

  6. Technology development of nuclear material safeguards for DUPIC fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jong Sook; Kim, Ho Dong; Kang, Hee Young; Lee, Young Gil; Byeon, Kee Ho; Park, Young Soo; Cha, Hong Ryul; Park, Ho Joon; Lee, Byung Doo; Chung, Sang Tae; Choi, Hyung Rae; Park, Hyun Soo

    1997-07-01

    During the second phase of research and development program conducted from 1993 to 1996, nuclear material safeguards studies system were performed on the technology development of DUPIC safeguards system such as nuclear material measurement in bulk form and product form, DUPIC fuel reactivity measurement, near-real-time accountancy, and containment and surveillance system for effective and efficient implementation of domestic and international safeguards obligation. By securing in advance a optimized safeguards system with domestically developed hardware and software, it will contribute not only to the effective implementation of DUPIC safeguards, but also to enhance the international confidence build-up in peaceful use of spent fuel material. (author). 27 refs., 13 tabs., 89 figs.

  7. Development of decommissioning management system for nuclear fuel cycle facilities (DECMAN)

    Energy Technology Data Exchange (ETDEWEB)

    Ogawa, Ryuichirou; Ishijima, Noboru; Tanimoto, Ken-ichi [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1999-04-01

    In making a plan of decommissioning of nuclear fuel facilities, it is important to optimize the plan on the standpoint of a few viewpoints, that is, the amount of working days, workers, radioactive waste, exposure dose of worker, and cost (they are called evaluation indexes). In the midst of decommissioning, the decommissioning plan would be modified suitably to optimize the evaluation indexes adjusting to progress of the decommissioning. The decommissioning management code (DECMAN), that is support system on computer, has been developed to assist the decommissioning planning. The system calculates the evaluation indexes quantitatively. The system consists of three fundamental codes, facility information database code, technical know-how database code and index evaluation code, they are composed using Oracle' database and 'G2' expert system. The functions of the system are as follows. (1) Facility information database code. Information of decommissioning facility and its rooms, machines and pipes in the code. (2) Technical know-how database code. Technical Information of tools to use in decommissioning work, cutting, dose measure, and decontamination are there. (3) Index evaluation code. User build decommissioning program using above two database codes. The code evaluates five indexes, the amount of working days, workers, radioactive waste, exposure dose of worker, and cost, on planning decommissioning program. Results of calculation are shown in table, chart, and etc. (author)

  8. Nuclear fuel cycle assessment of India: A technical study for U.S.-India cooperation

    Science.gov (United States)

    Krishna, Taraknath Woddi Venkat

    The recent civil nuclear cooperation proposed by the Bush Administration and the Government of India has heightened the necessity of assessing India's nuclear fuel cycle inclusive of nuclear materials and facilities. This agreement proposes to change the long-standing U.S. policy of preventing the spread of nuclear weapons by denying nuclear technology transfer to non-NPT signatory states. The nuclear tests in 1998 have convinced the world community that India would never relinquish its nuclear arsenal. This has driven the desire to engage India through civilian nuclear cooperation. The cornerstone of any civilian nuclear technological support necessitates the separation of military and civilian facilities. A complete nuclear fuel cycle assessment of India emphasizes the entwinment of the military and civilian facilities and would aid in moving forward with the separation plan. To estimate the existing uranium reserves in India, a complete historical assessment of ore production, conversion, and processing capabilities was performed using open source information and compared to independent reports. Nuclear energy and plutonium production (reactor- and weapons-grade) was simulated using declared capacity factors and modern simulation tools. The three-stage nuclear power program entities and all the components of civilian and military significance were assembled into a flowsheet to allow for a macroscopic vision of the Indian fuel cycle. A detailed view of the nuclear fuel cycle opens avenues for technological collaboration. The fuel cycle that grows from this study exploits domestic thorium reserves with advanced international technology and optimized for the existing system. To utilize any appreciable fraction of the world's supply of thorium, nuclear breeding is necessary. The two known possibilities for production of more fissionable material in the reactor than is consumed as fuel are fast breeders or thermal breeders. This dissertation analyzes a thermal

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

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Lineberry, M.J. [Argonne National Lab., Idaho Falls, ID (United States). Technology Development Div.

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Lineberry, M.J. [Argonne National Lab., Idaho Falls, ID (United States). Technology Development Div.

    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.

  11. Modeling and analysis of tritium dynamics in a DT fusion fuel cycle

    Science.gov (United States)

    Kuan, William

    1998-11-01

    A number of crucial design issues have a profound effect on the dynamics of the tritium fuel cycle in a DT fusion reactor, where the development of appropriate solutions to these issues is of particular importance to the introduction of fusion as a commercial system. Such tritium-related issues can be classified according to their operational, safety, and economic impact to the operation of the reactor during its lifetime. Given such key design issues inherent in next generation fusion devices using the DT fuel cycle development of appropriate models can then lead to optimized designs of the fusion fuel cycle for different types of DT fusion reactors. In this work, two different types of modeling approaches are developed and their application to solving key tritium issues presented. For the first approach, time-dependent inventories, concentrations, and flow rates characterizing the main subsystems of the fuel cycle are simulated with a new dynamic modular model of a fusion reactor's fuel cycle, named X-TRUFFLES (X-Windows TRitiUm Fusion Fuel cycLE dynamic Simulation). The complex dynamic behavior of the recycled fuel within each of the modeled subsystems is investigated using this new integrated model for different reactor scenarios and design approaches. Results for a proposed fuel cycle design taking into account current technologies are presented, including sensitivity studies. Ways to minimize the tritium inventory are also assessed by examining various design options that could be used to minimize local and global tritium inventories. The second modeling approach involves an analytical model to be used for the calculation of the required tritium breeding ratio, i.e., a primary design issue which relates directly to the feasibility and economics of DT fusion systems. A time-integrated global tritium balance scheme is developed and appropriate analytical expressions are derived for tritium self-sufficiency relevant parameters. The easy exploration of the large

  12. Prediction Method of the Fuel Consumption of Wheel Loaders in the V-Type Loading Cycle

    Directory of Open Access Journals (Sweden)

    Wenxing Ma

    2015-01-01

    Full Text Available Wheel loaders in the V-type loading cycle are characterized by complicated loading conditions, nonlinear power-train system, and time-variable engine power distribution. Therefore, it is difficult to predict the fuel consumption of wheel loaders in the V-type loading cycle. The static matching methods cannot provide fuel consumption prediction for the loading cycle. In this paper, the prediction method and model of the fuel consumption for wheel loaders in the V-type loading cycle were proposed. Firstly, the hydraulic system data were tested when a wheel loader loaded three different materials in a typical V-type loading cycle. Secondly, the tested data were filtered by the 8th-order Butterworth filter and the dimensionless power deduction equations of hydraulic power system for loading three different materials were obtained by Gaussian and linear fitting based on the filtered data in the loading cycle. Finally, fuel consumption was obtained with the compiling dynamic calculation program as well as input parameters of tested vehicle speed, throttle parameter, and the dimensionless equation. The simulation results agreed well with experiment results. Dynamic calculation program is applicable to calculate loading economy and can provide academic guidance for wheel loader’s design and optimization.

  13. ASSERT-PV 3.2: Advanced subchannel thermalhydraulics code for CANDU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Y.F., E-mail: raoy@aecl.ca; Cheng, Z., E-mail: chengz@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca; Nava-Dominguez, A., E-mail: navadoma@aecl.ca

    2014-08-15

    Highlights: • Introduction to a new version of the Canadian subchannel code, ASSERT-PV 3.2. • Enhanced models for flow-distribution, CHF and post-dryout heat transfer prediction. • Model changes focused on unique features of horizontal CANDU bundles. • Detailed description of model changes for all major thermalhydraulics models. • Discussion on rationale and limitation of the model changes. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The most recent release version, ASSERT-PV 3.2 has enhanced phenomenon models for improved predictions of flow distribution, dryout power and CHF location, and post-dryout (PDO) sheath temperature in horizontal CANDU fuel bundles. The focus of the improvements is mainly on modeling considerations for the unique features of CANDU bundles such as horizontal flows, small pitch to diameter ratios, high mass fluxes, and mixed and irregular subchannel geometries, compared to PWR/BWR fuel assemblies. This paper provides a general introduction to ASSERT-PV 3.2, and describes the model changes or additions in the new version to improve predictions of flow distribution, dryout power and CHF location, and PDO sheath temperatures in CANDU fuel bundles.

  14. Improved solidification influence modelling for Eulerian fuel-coolant interaction codes

    Energy Technology Data Exchange (ETDEWEB)

    Ursic, Mitja, E-mail: mitja.ursic@ijs.s [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Leskovar, Matjaz; Mavko, Borut [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2011-04-15

    Steam explosion experiments revealed important differences in the efficiency between simulant alumina and oxidic corium melts. The experimentally observed differences are importantly attributed to the differences in the melt droplets solidification and void production, which are limiting phenomena in the steam explosion process and have to be adequately modelled in fuel-coolant interaction codes. This article focuses on the modelling of the solidification effect. An improved solidification influence modelling approach for Eulerian fuel-coolant interaction codes was developed and is presented herein. The solidification influence modelling in fuel-coolant interaction codes is strongly related to the modelling of the temperature profile and the mechanical effect of the crust on the fragmentation process. Therefore the first objective was to introduce an improved temperature profile modelling and a fragmentation criterion for partly solidified droplets. The fragmentation criterion was based on the established modified Weber number, which considers the crust stiffness as a stabilizing force acting to retain the crust under presence of the hydrodynamic forces. The modified Weber number was validated on experimental data. The application of the developed improved solidification influence modelling enables an improved determination of the melt droplet mass, which can be efficiently involved in the fine fragmentation during the steam explosion process. Additionally, also the void production modelling is improved, because it is strongly related to the temperature profile modelling in the frame of the solidification influence modelling. Therefore the second objective was to enable an improved solidification influence modelling in codes with an Eulerian formulation of the droplet field. Two additional transported model parameters based on the most important droplets features regarding the fuel-coolant interaction behaviour, were derived. First, the crust stiffness was

  15. Interim storage of power reactor spent nuclear fuel (SNF) and its potential application to SNF separations and closed fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Levy, Salomon, E-mail: slevy112@aol.com

    2009-10-15

    Interim, centralized, engineered (dry cask) storage facilities for USA light water power reactor spent nuclear fuel (SNF) should be implemented to complement and to offer much needed flexibility while the Nuclear Regulatory Commission is funded to complete its evaluation of the Yucca Mountain License and to subject it to public hearings. The interim sites should use the credo reproduced in Table 1 [Bunn, M., 2001. Interim Storage of Spent Nuclear Fuel. Harvard University and University of Tokyo] and involve both the industry and government. The sites will help settle the 50 pending lawsuits against the government and the $11 billion of potential additional liabilities for SNF delay damages if Yucca Mountain does not being operation in 2020 [DOE, 2008a. Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel from Decommissioned Nuclear Power Stations (December)]. Under the developing consensus to proceed with closed fuel cycles, it will be necessary to develop SNF separation facilities with stringent requirements upon separation processes and upon generation of only highly resistant waste forms. The location of such facilities at the interim storage sites would offer great benefits to those sites and assure their long term viability by returning them to their original status. The switch from once-through to closed fuel cycle will require extensive time and development work as illustrated in 'The Path to Sustainable Nuclear Energy' [DOE, 2005. The Path to Sustainable Nuclear Energy. Basic and Applied Research Opportunities for Advanced Fuel Cycles. DOE (September)]. A carefully crafted long term program, funded for at least 5 years, managed by a strong joint government-industry team, and subjected to regular independent reviews should be considered to assure the program stability and success. The new uncertainty about Yucca Mountain role raises two key issues: (a) what to do with the weapons and other high level government

  16. Accounting for water formation from hydrocarbon fuel combustion in life cycle analyses

    Science.gov (United States)

    Belmont, E. L.; Davidson, F. T.; Glazer, Y. R.; Beagle, E. A.; Webber, M. E.

    2017-09-01

    Hydrocarbon fuel production and utilization are considered water intensive processes due to the high volumes of water used in source development and fuel processing. At the same time, there is significant water formed during combustion. However, this water is not currently widely harvested at the site of production. Instead, it is added to the hydrologic cycle, often in a different location from the fuel production site. This study quantifies the water formed from combustion of these fuels and analyzes the magnitudes of formation in the context of other hydrologic sources and sinks in order to facilitate future assessments of water harvesting technology and/or atmospheric impacts of combustion. Annual water formation from stoichiometric combustion of hydrocarbon fuels, including natural gas, oil- and natural gas liquid-derived products, and coal, in the United States and worldwide are presented and compared with quantities of water sequestered, evaporated, and stored in the atmosphere. Water production factors in terms of mass and energy of fuel consumed, WPFm and WPFe, respectively, are defined for the comparison of fuels and incorporation into future life cycle analyses (LCAs). Results show that water formation from combustion has increased worldwide from 2005 to 2015, with the largest increase coming from growth in combustion of natural gas. Water formation from combustion of hydrocarbon fuels equals or exceeds water sequestered from the hydrologic cycle through deep well injection in the US annually. Overall, water formation is deemed significant enough to warrant consideration by LCAs of water intensity in fuel production and use, and should be included in future analyses.

  17. Quantitative Analysis of the Civilian Bilateral Cooperation in Front-End of the Nuclear Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Viet Phuong; Yim, Man-Sung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-05-15

    A substantial part of such cooperation is related to the front-end of the nuclear fuel cycle, which encompasses the processes that help manufacturing nuclear fuel, including mining and milling of natural uranium, refining and chemical conversion, enrichment (in case of fuels for Pressurized Water Reactor PWR), and fuel fabrication. Traditionally, the supply of natural uranium was dominated by Canada and Australia, whereas enrichment services have been mostly provided by companies from Western states or Russia, which are also the main customers of such services. However, Kazakhstan and African countries like Niger, Namibia, and Malawi have emerged as important suppliers in the international uranium market and recent forecasts show that China will soon become a major player in the front-end market as both consumer and service provider. In this paper, the correlation between bilateral civil nuclear cooperation in front-end of the nuclear fuel cycle and the political and economic relationship among countries was examined through a dataset of bilateral nuclear cooperation in the post-Cold War era, from 1990 to 2011. Such finding has implication on not only the nonproliferation research but also the necessary reinforcement of export control regimes like such as the Nuclear Suppliers Group. Further improvement of this dataset and the regression method are also needed in order to increase the robustness of the findings as well as to cover the whole scope of the nuclear fuel cycle, including both front-end and back-end activities.

  18. Summary of Off-Normal Events in US Fuel Cycle Facilities for AFCI Applications

    Energy Technology Data Exchange (ETDEWEB)

    L. C. Cadwallader; S. J. Piet; S. O. Sheetz; D. H. McGuire; W. B. Boore

    2005-09-01

    This report is a collection and review of system operation and failure experiences for facilities comprising the fission reactor fuel cycle, with the exception of reactor operations. This report includes mines, mills, conversion plants, enrichment plants, fuel fabrication plants, transportation of fuel materials between these centers, and waste storage facilities. Some of the facilities discussed are no longer operating; others continue to produce fuel for the commercial fission power plant industry. Some of the facilities discussed have been part of the military’s nuclear effort; these are included when the processes used are similar to those used for commercial nuclear power. When reading compilations of incidents and accidents, after repeated entries it is natural to form an opinion that there exists nothing but accidents. For this reason, production or throughput values are described when available. These adverse operating experiences are compiled to support the design and decisions needed for the Advanced Fuel Cycle Initiative (AFCI). The AFCI is to weigh options for a new fission reactor fuel cycle that is efficient, safe, and productive for US energy security.

  19. Energy and emission benefits of alternative transportation liquid fuels derived from switchgrass: a fuel life cycle assessment.

    Science.gov (United States)

    Wu, May; Wu, Ye; Wang, Michael

    2006-01-01

    We conducted a mobility chains, or well-to-wheels (WTW), analysis to assess the energy and emission benefits of cellulosic biomass for the U.S. transportation sector in the years 2015-2030. We estimated the life-cycle energy consumption and emissions associated with biofuel production and use in light-duty vehicle (LDV) technologies by using the Greenhouse gases, Regulated Emissions, and Energy use in Transportation (GREET) model. Analysis of biofuel production was based on ASPEN Plus model simulation of an advanced fermentation process to produce fuel ethanol/protein, a thermochemical process to produce Fischer-Tropsch diesel (FTD) and dimethyl ether (DME), and a combined heat and power plant to co-produce steam and electricity. Our study revealed that cellulosic biofuels as E85 (mixture of 85% ethanol and 15% gasoline by volume), FTD, and DME offer substantial savings in petroleum (66-93%) and fossil energy (65-88%) consumption on a per-mile basis. Decreased fossil fuel use translates to 82-87% reductions in greenhouse gas emissions across all unblended cellulosic biofuels. In urban areas, our study shows net reductions for almost all criteria pollutants, with the exception of carbon monoxide (unchanged), for each of the biofuel production option examined. Conventional and hybrid electric vehicles, when fueled with E85, could reduce total sulfur oxide (SO(x)) emissions to 39-43% of those generated by vehicles fueled with gasoline. By using bio-FTD and bio-DME in place of diesel, SO(x) emissions are reduced to 46-58% of those generated by diesel-fueled vehicles. Six different fuel production options were compared. This study strongly suggests that integrated heat and power co-generation by means of gas turbine combined cycle is a crucial factor in the energy savings and emission reductions.

  20. LIFE CYCLE DESIGN OF A FUEL TANK SYSTEM

    Science.gov (United States)

    This life cycle design (LCD) project was a collaborative effort between the National Pollution Prevention Center at the University of Michigan, General Motors (GM), and the U.S. Environmental Protection Agency (EPA). The primary objective of this project was to apply life cyc...

  1. TPP Fuel Characteristic – Objective Index of Its Cycle Economy

    Directory of Open Access Journals (Sweden)

    A. E. Piir

    2011-01-01

    Full Text Available The paper presents a methodology and an example how to calculate thermo-dynamically  justified specific consumer of reference fuel at combined generation of heat and electric power. The methodology is based on an analytical generalization of mode diagrams of the turbine which installed at the TPP.While calculating self-cost the paper proposes to exclude a conditional division of the equipment into heat- or electric power generating one and heat tariff calculation is carried out on the basis of a market price for electric power. 

  2. Characterizing model uncertainties in the life cycle of lignocellulose-based ethanol fuels.

    Science.gov (United States)

    Spatari, Sabrina; MacLean, Heather L

    2010-11-15

    Renewable and low carbon fuel standards being developed at federal and state levels require an estimation of the life cycle carbon intensity (LCCI) of candidate fuels that can substitute for gasoline, such as second generation bioethanol. Estimating the LCCI of such fuels with a high degree of confidence requires the use of probabilistic methods to account for known sources of uncertainty. We construct life cycle models for the bioconversion of agricultural residue (corn stover) and energy crops (switchgrass) and explicitly examine uncertainty using Monte Carlo simulation. Using statistical methods to identify significant model variables from public data sets and Aspen Plus chemical process models,we estimate stochastic life cycle greenhouse gas (GHG) emissions for the two feedstocks combined with two promising fuel conversion technologies. The approach can be generalized to other biofuel systems. Our results show potentially high and uncertain GHG emissions for switchgrass-ethanol due to uncertain CO₂ flux from land use change and N₂O flux from N fertilizer. However, corn stover-ethanol,with its low-in-magnitude, tight-in-spread LCCI distribution, shows considerable promise for reducing life cycle GHG emissions relative to gasoline and corn-ethanol. Coproducts are important for reducing the LCCI of all ethanol fuels we examine.

  3. Study of visualized simulation and analysis of nuclear fuel cycle system based on multilevel flow model

    Institute of Scientific and Technical Information of China (English)

    LIU Jing-Quan; YOSHIKAWA Hidekazu; ZHOU Yang-Ping

    2005-01-01

    Complex energy and environment system, especially nuclear fuel cycle system recently raised social concerns about the issues of economic competitiveness, environmental effect and nuclear proliferation. Only under the condition that those conflicting issues are gotten a consensus between stakeholders with different knowledge background, can nuclear power industry be continuingly developed. In this paper, a new analysis platform has been developed to help stakeholders to recognize and analyze various socio-technical issues in the nuclear fuel cycle system based on the functional modeling method named Multilevel Flow Models (MFM) according to the cognition theory of human being. Its character is that MFM models define a set of mass, energy and information flow structures on multiple levels of abstraction to describe the functional structure of a process system and its graphical symbol representation and the means-end and part-whole hierarchical flow structure to make the represented process easy to be understood. Based upon this methodology, a micro-process and a macro-process of nuclear fuel cycle system were selected to be simulated and some analysis processes such as economics analysis, environmental analysis and energy balance analysis related to those flows were also integrated to help stakeholders to understand the process of decision-making with the introduction of some new functions for the improved Multilevel Flow Models Studio, and finally the simple simulation such as spent fuel management process simulation and money flow of nuclear fuel cycle and its levelised cost analysis will be represented as feasible examples.

  4. U.S. -- EC fuel cycle study: Background document to the approach and issues

    Energy Technology Data Exchange (ETDEWEB)

    Cantor, Robin; Russell, Lee; Krupnick, Alan; Smith, Hilary; Schaffhauser, Jr., A.; Barnthouse, Larry; Cada, Glen; Kroodsma, Roger; Turner, Robb; Easterly, Clay; Jones, Troyce; Burtraw, Dallas; Harrington, Winston; Freeman, A. Myrick

    1992-11-01

    In February 1991, DOE and the Commission of the European Communities (EC), signed a joint statement regarding the external costs of fuel cycles. This 18-month agreement committed their respective organizations to develop a comparative analytical methodology and to develop the best range of estimates of external costs from secondary sources'' for eight fuel cycles and four conservation options. In our study, a fuel cycle is defined as the series of physical and chemical processes and activities that are required to generate electricity from a specific fuel or resource. This foundation phase of the study is primarily limited to developing and demonstrating methods for estimating impacts and their monetized value, what we term damages'' or benefits,'' leaving aside the extent to which such damages have been internalized. However, Appendix C provides the conceptual framework for evaluating the extent of internalization. This report is a background document to introduce the study approach and to discuss the major conceptual and practical issues entailed by the incremental damage problem. As a background document, the report seeks to communicate an overview of the study and the important methodological choices that were made to conduct the research. In successive sections of the report, the methodological tools used in the study are discussed; the ecological and health impacts are reviewed using the coal fuel cycle as a reference case; and, in the final chapter, the methods for valuing impacts are detailed.

  5. A multi-platform linking code for fuel burnup and radiotoxicity analysis

    Science.gov (United States)

    Cunha, R.; Pereira, C.; Veloso, M. A. F.; Cardoso, F.; Costa, A. L.

    2014-02-01

    A linking code between ORIGEN2.1 and MCNP has been developed at the Departamento de Engenharia Nuclear/UFMG to calculate coupled neutronic/isotopic results for nuclear systems and to produce a large number of criticality, burnup and radiotoxicity results. In its previous version, it evaluated the isotopic composition evolution in a Heat Pipe Power System model as well as the radiotoxicity and radioactivity during lifetime cycles. In the new version, the code presents features such as multi-platform execution and automatic results analysis. Improvements made in the code allow it to perform simulations in a simpler and faster way without compromising accuracy. Initially, the code generates a new input for MCNP based on the decisions of the user. After that, MCNP is run and data, such as recoverable energy per prompt fission neutron, reaction rates and keff, are automatically extracted from the output and used to calculate neutron flux and cross sections. These data are then used to construct new ORIGEN inputs, one for each cell in the core. Each new input is run on ORIGEN and generates outputs that represent the complete isotopic composition of the core on that time step. The results show good agreement between GB (Coupled Neutronic/Isotopic code) and Monteburns (Automated, Multi-Step Monte Carlo Burnup Code System), developed by the Los Alamos National Laboratory.

  6. New features and applications of PRESTO, a computer code for the performance of regenerative, superheated steam turbine cycles

    Science.gov (United States)

    Choo, Y. K.; Staiger, P. J.

    1982-01-01

    The code was designed to analyze performance at valves-wide-open design flow. The code can model conventional steam cycles as well as cycles that include such special features as process steam extraction and induction and feedwater heating by external heat sources. Convenience features and extensions to the special features were incorporated into the PRESTO code. The features are described, and detailed examples illustrating the use of both the original and the special features are given.

  7. REVA Advanced Fuel Design and Codes and Methods - Increasing Reliability, Operating Margin and Efficiency in Operation

    Energy Technology Data Exchange (ETDEWEB)

    Frichet, A.; Mollard, P.; Gentet, G.; Lippert, H. J.; Curva-Tivig, F.; Cole, S.; Garner, N.

    2014-07-01

    Since three decades, AREVA has been incrementally implementing upgrades in the BWR and PWR Fuel design and codes and methods leading to an ever greater fuel efficiency and easier licensing. For PWRs, AREVA is implementing upgraded versions of its HTP{sup T}M and AFA 3G technologies called HTP{sup T}M-I and AFA3G-I. These fuel assemblies feature improved robustness and dimensional stability through the ultimate optimization of their hold down system, the use of Q12, the AREVA advanced quaternary alloy for guide tube, the increase in their wall thickness and the stiffening of the spacer to guide tube connection. But an even bigger step forward has been achieved a s AREVA has successfully developed and introduces to the market the GAIA product which maintains the resistance to grid to rod fretting (GTRF) of the HTP{sup T}M product while providing addition al thermal-hydraulic margin and high resistance to Fuel Assembly bow. (Author)

  8. Influence of driving cycles on exhaust emissions and fuel consumption of gasoline passenger car in Bangkok.

    Science.gov (United States)

    Nutramon, Tamsanya; Supachart, Chungpaibulpatana

    2009-01-01

    The influence of different driving cycles on their exhaust emissions and fuel consumption rate of gasoline passenger car was investigated in Bangkok based on the actual measurements obtained from a test vehicle driving on a standard chassis dynamometer. A newly established Bangkok driving cycle (BDC) and the European driving cycle (EDC) which is presently adopted as the legislative cycle for testing automobiles registered in Thailand were used. The newly developed BDC is constructed using the driving characteristic data obtained from the real on-road driving tests along selected traffic routes. A method for selecting appropriate road routes for real driving tests is also introduced. Variations of keyed driving parameters of BDC with different driving cycles were discussed. The results showed that the HC and CO emission factors of BDC are almost two and four times greater than those of EDC, respectively. Although the difference in the NOx emission factor is small, the value from BDC is still greater than that of EDC by 10%. Under BDC, the test vehicle consumes fuel about 25% more than it does under EDC. All these differences are mainly attributed to the greater proportion of idle periods and higher fluctuations of vehicle speed in the BDC cycle. This result indicated that the exhausted emissions and fuel consumption of vehicles obtained from tests under the legislative modal-type driving cycle (EDC) are significantly different from those actually produced under real traffic conditions especially during peak periods.

  9. Influence of driving cycles on exhaust emissions and fuel consumption of gasoline passenger car in Bangkok

    Institute of Scientific and Technical Information of China (English)

    NUTRAMON Tamsanya; SUPACHART Chungpaibulpatana

    2009-01-01

    The influence of different driving cycles on their exhaust emissions and fuel consumption rate of gasoline passenger car was investigated in Bangkok based on the actual measurements obtained from a test vehicle driving on a standard chassis dynamometer. A newly established Bangkok driving cycle (BDC) and the European driving cycle (EDC) which is presently adopted as the legislative cycle for testing automobiles registered in Thailand were used. The newly developed BDC is constructed using the driving characteristic data obtained from the real on-road driving tests along selected traffic routes. A method for selecting appropriate road routes for real driving tests is also introduced. Variations of keyed driving parameters of BDC with different driving cycles were discussed. The results showed that the HC and CO emission factors of BDC are almost two and four times greater than those of EDC, respectively. Although the difference in the NOx emission factor is small, the value from BDC is still greater than that of EDC by 10%. Under BDC, the test vehicle consumes fuel about 25% more than it does under EDC. All these differences are mainly attributed to the greater proportion of idle periods and higher fluctuations of vehicle speed in the BDC cycle. This result indicated that the exhausted emissions and fuel consumption of vehicles obtained from tests under the legislative modal-type driving cycle (EDC) are significantly different from those actually produced under real traffic conditions especially during peak periods.

  10. Graphical and tabular summaries of decay characteristics for once-through PWR, LMFBR, and FFTF fuel cycle materials. [Spent fuel, high-level waste fuel can scrap

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A.G.; Liberman, M.S.; Morrison, G.W.

    1982-01-01

    Based on the results of ORIGEN2 and a newly developed code called ORMANG, graphical and summary tabular characteristics of spent fuel, high-level waste, and fuel assembly structural material (cladding) waste are presented for a generic pressurized-water reactor (PWR), a liquid-metal fast breeder reactor (LMFBR), and the Fast Flux Test Facility (FFTF). The characteristics include radioactivity, thermal power, and toxicity (water dilution volume). Given are graphs and summary tables containing characteristic totals and the principal nuclide contributors as well as graphs comparing the three reactors for a single material and the three materials for a single reactor.

  11. International Source Book: Nuclear Fuel Cycle Research and Development Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, K. M.; Lakey, L. T.

    1982-11-01

    This document starts with an overview that summarizes nuclear power policies and waste management activities for nations with significant commercial nuclear fuel cycle activities either under way or planned. A more detailed program summary is then included for each country or international agency conducting nuclear fuel cycle and waste management research and development. This second volume includes the program summaries of those countries listed alphabetically from Japan to Yugoslavia. Information on international agencies and associations, particularly the IAEA, NEA, and CEC, is provided also.

  12. Recovery of Information from the Fast Flux Test Facility for the Advanced Fuel Cycle Initiative

    Energy Technology Data Exchange (ETDEWEB)

    Nielsen, Deborah L.; Makenas, Bruce J.; Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.

    2009-09-30

    The Fast Flux Test Facility is the most recent Liquid Metal Reactor to operate in the United States. Information from the design, construction, and operation of this reactor was at risk as the facilities associated with the reactor are being shut down. The Advanced Fuel Cycle Initiative is a program managed by the Office of Nuclear Energy of the U.S. Department of Energy with a mission to develop new fuel cycle technologies to support both current and advanced reactors. Securing and preserving the knowledge gained from operation and testing in the Fast Flux Test Facility is an important part of the Knowledge Preservation activity in this program.

  13. DESIGN OF COMBINED CYCLE GENERATION SYSTEM WITH HIGH TEMPERATURE FUEL CELL AND STEAM TURBINE

    Institute of Scientific and Technical Information of China (English)

    Yu Lijun; Yuan Junqi; Cao Guangyi

    2003-01-01

    For environment protection and high efficiency, development of new concept power plant has been required in China. The fuel cell is expected to be used in a power plant as a centralized power station or distributed power plant. It is a chemical power generation device that converts the energy of a chemical reaction directly into electrical energy and not limited by Carnot cycle efficiency. The molten carbonate fuel cell (MCFC) power plant has several attractive features I.e. High efficiency and lower emission of Nox and Sox. A combined cycle generation system with MCFC and steam turbine is designed. Its net electrical efficiency LHV is about 55%.

  14. Identifying codes and locating-dominating sets on paths and cycles

    CERN Document Server

    Chen, Chunxia; Miao, Zhengke

    2009-01-01

    Let $G=(V,E)$ be a graph and let $r\\ge 1$ be an integer. For a set $D \\subseteq V$, define $N_r[x] = \\{y \\in V: d(x, y) \\leq r\\}$ and $D_r(x) = N_r[x] \\cap D$, where $d(x,y)$ denotes the number of edges in any shortest path between $x$ and $y$. $D$ is known as an $r$-identifying code ($r$-locating-dominating set, respectively), if for all vertices $x\\in V$ ($x \\in V\\backslash D$, respectively), $D_r(x)$ are all nonempty and different. In this paper, we provide complete results for $r$-identifying codes in paths and odd cycles; we also give complete results for 2-locating-dominating sets in cycles.

  15. Solid Oxide Fuel Cell/Gas Turbine Hybrid Cycle Technology for Auxiliary Aerospace Power

    Science.gov (United States)

    Steffen, Christopher J., Jr.; Freeh, Joshua E.; Larosiliere, Louis M.

    2005-01-01

    A notional 440 kW auxiliary power unit has been developed for 300 passenger commercial transport aircraft in 2015AD. A hybrid engine using solid-oxide fuel cell stacks and a gas turbine bottoming cycle has been considered. Steady-state performance analysis during cruise operation has been presented. Trades between performance efficiency and system mass were conducted with system specific energy as the discriminator. Fuel cell performance was examined with an area specific resistance. The ratio of fuel cell versus turbine power was explored through variable fuel utilization. Area specific resistance, fuel utilization, and mission length had interacting effects upon system specific energy. During cruise operation, the simple cycle fuel cell/gas turbine hybrid was not able to outperform current turbine-driven generators for system specific energy, despite a significant improvement in system efficiency. This was due in part to the increased mass of the hybrid engine, and the increased water flow required for on-board fuel reformation. Two planar, anode-supported cell design concepts were considered. Designs that seek to minimize the metallic interconnect layer mass were seen to have a large effect upon the system mass estimates.

  16. Quantifying variability in life cycle greenhouse gas inventories of alternative middle distillate transportation fuels.

    Science.gov (United States)

    Stratton, Russell W; Wong, Hsin Min; Hileman, James I

    2011-05-15

    The presence of variability in life cycle analysis (LCA) is inherent due to both inexact LCA procedures and variation of numerical inputs. Variability in LCA needs to be clearly distinguished from uncertainty. This paper uses specific examples from the production of diesel and jet fuels from 14 different feedstocks to demonstrate general trends in the types and magnitudes of variability present in life cycle greenhouse gas (LC-GHG) inventories of middle distillate fuels. Sources of variability have been categorized as pathway specific, coproduct usage and allocation, and land use change. The results of this research demonstrate that subjective choices such as coproduct usage and allocation methodology can be more important sources of variability in the LC-GHG inventory of a fuel option than the process and energy use of fuel production. Through the application of a consistent analysis methodology across all fuel options, the influence of these subjective biases is minimized, and the LC-GHG inventories for each feedstock-to-fuel option can be effectively compared and discussed. By considering the types and magnitudes of variability across multiple fuel pathways, it is evident that LCA results should be presented as a range instead of a point value. The policy implications of this are discussed.

  17. Cycle analysis of an integrated solid oxide fuel cell and recuperative gas turbine with an air reheating system

    Science.gov (United States)

    Zhang, Xiongwen; Li, Jun; Li, Guojun; Feng, Zhenping

    Cycle simulation and analysis for two kinds of SOFC/GT hybrid systems were conducted with the help of the simulation tool: Aspen Custom Modeler. Two cycle schemes of recuperative heat exchanger (RHE) and exhaust gas recirculated (EGR) were described according to the air reheating method. The system performance with operating pressure, turbine inlet temperature and fuel cell load were studied based on the simulation results. Then the effects of oxygen utilization, fuel utilization, operating temperature and efficiencies of the gas turbine components on the system performance of the RHE cycle and the EGR cycle were discussed in detail. Simulation results indicated that the system optimum efficiency for the EGR air reheating cycle scheme was higher than that of the RHE cycle system. A higher pressure ratio would be available for the EGR cycle system in comparison with the RHE cycle. It was found that increasing fuel utilization or oxygen utilization would decrease fuel cell efficiency but improve the system efficiency for both of the RHE and EGR cycles. The efficiency of the RHE cycle hybrid system decreased as the fuel cell air inlet temperature increased. However, the system efficiency of EGR cycle increased with fuel cell air inlet temperature. The effect of turbine efficiency on the system efficiency was more obvious than the effect of the compressor and recuperator efficiencies among the gas turbine components. It was also indicated that improving the gas turbine component efficiencies for the RHE cycle increased system efficiency higher than that for the EGR cycle.

  18. Nuclear fuel cycle and sustainable development: strategies for the future

    Energy Technology Data Exchange (ETDEWEB)

    Bouchard, J

    2004-07-01

    In this presentation, the author aims to define the major role of the nuclear energy in the future, according a sustainable development scenario. The today aging park and the new Generation IV technologies are presented. The transition scenario from Pu mono-recycling in PWRs to actinide global recycling in fast neutron Gen IV systems is also developed. Closed cycles and fast reactors appear as the appropriate answer to sustainable objectives in a vision of a large expansion. (A.L.B.)

  19. A Review of RedOx Cycling of Solid Oxide Fuel Cells Anode

    Directory of Open Access Journals (Sweden)

    Jan Van herle

    2012-08-01

    Full Text Available Solid oxide fuel cells are able to convert fuels, including hydrocarbons, to electricity with an unbeatable efficiency even for small systems. One of the main limitations for long-term utilization is the reduction-oxidation cycling (RedOx cycles of the nickel-based anodes. This paper will review the effects and parameters influencing RedOx cycles of the Ni-ceramic anode. Second, solutions for RedOx instability are reviewed in the patent and open scientific literature. The solutions are described from the point of view of the system, stack design, cell design, new materials and microstructure optimization. Finally, a brief synthesis on RedOx cycling of Ni-based anode supports for standard and optimized microstructures is depicted.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. 40 CFR 600.207-08 - Calculation and use of vehicle-specific 5-cycle-based fuel economy values for vehicle...

    Science.gov (United States)

    2010-07-01

    .... (b) If only one equivalent petroleum-based fuel economy value exists for an electric configuration...-cycle fuel economy for that configuration. (c) If more than one equivalent petroleum-based 5-cycle fuel...-specific 5-cycle-based fuel economy values for vehicle configurations. 600.207-08 Section...

  2. User Guide for VISION 3.4.7 (Verifiable Fuel Cycle Simulation) Model

    Energy Technology Data Exchange (ETDEWEB)

    Jacob J. Jacobson; Robert F. Jeffers; Gretchen E. Matthern; Steven J. Piet; Wendell D. Hintze

    2011-07-01

    The purpose of this document is to provide a guide for using the current version of the Verifiable Fuel Cycle Simulation (VISION) model. This is a complex model with many parameters and options; the user is strongly encouraged to read this user guide before attempting to run the model. This model is an R&D work in progress and may contain errors and omissions. It is based upon numerous assumptions. This model is intended to assist in evaluating 'what if' scenarios and in comparing fuel, reactor, and fuel processing alternatives at a systems level. The model is not intended as a tool for process flow and design modeling of specific facilities nor for tracking individual units of fuel or other material through the system. The model is intended to examine the interactions among the components of a fuel system as a function of time varying system parameters; this model represents a dynamic rather than steady-state approximation of the nuclear fuel system. VISION models the nuclear cycle at the system level, not individual facilities, e.g., 'reactor types' not individual reactors and 'separation types' not individual separation plants. Natural uranium can be enriched, which produces enriched uranium, which goes into fuel fabrication, and depleted uranium (DU), which goes into storage. Fuel is transformed (transmuted) in reactors and then goes into a storage buffer. Used fuel can be pulled from storage into either separation or disposal. If sent to separations, fuel is transformed (partitioned) into fuel products, recovered uranium, and various categories of waste. Recycled material is stored until used by its assigned reactor type. VISION is comprised of several Microsoft Excel input files, a Powersim Studio core, and several Microsoft Excel output files. All must be co-located in the same folder on a PC to function. You must use Powersim Studio 8 or better. We have tested VISION with the Studio 8 Expert, Executive, and Education versions

  3. Life cycle assessment of fuel ethanol derived from corn grain via dry milling.

    Science.gov (United States)

    Kim, Seungdo; Dale, Bruce E

    2008-08-01

    Life cycle analysis enables to investigate environmental performance of fuel ethanol used in an E10 fueled compact passenger vehicle. Ethanol is derived from corn grain via dry milling. This type of analysis is an important component for identifying practices that will help to ensure that a renewable fuel, such as ethanol, may be produced in a sustainable manner. Based on data from eight counties in seven Corn Belt states as corn farming sites, we show ethanol derived from corn grain as E10 fuel would reduce nonrenewable energy and greenhouse gas emissions, but would increase acidification, eutrophication and photochemical smog, compared to using gasoline as liquid fuel. The ethanol fuel systems considered in this study offer economic benefits, namely more money returned to society than the investment for producing ethanol. The environmental performance of ethanol fuel system varies significantly with corn farming sites because of different crop management practices, soil properties, and climatic conditions. The dominant factor determining most environmental impacts considered here (i.e., greenhouse gas emissions, acidification, eutrophication, and photochemical smog formation) is soil related nitrogen losses (e.g., N2O, NOx, and NO3-). The sources of soil nitrogen include nitrogen fertilizer, crop residues, and air deposition. Nitrogen fertilizer is probably the primary source. Simulations using an agro-ecosystem model predict that planting winter cover crops would reduce soil nitrogen losses and increase soil organic carbon levels, thereby greatly improving the environmental performance of the ethanol fuel system.

  4. External costs of the nuclear fuel cycle. A scoping study to determine the external costs of the Dutch nuclear fuel cycle in accordance with the EC/US methodology

    Energy Technology Data Exchange (ETDEWEB)

    Dodd, D.H.

    1995-10-01

    This report describes the results of a scoping study to estimate the external costs of the Dutch nuclear fuel cycle. This study was performed within the framework of the Commission of the European Community`s External Costs of Fuel Cycles project. The external costs of a fuel cycle are those costs which are excluded from the standard calculation of the cost of electricity. These costs are borne by society as a whole and include, in particular, the health and environmental costs which result from the operation of the facilities involved in a given fuel cycle. At present the uranium enrichment, electricity generation and interim storage stages of the nuclear fuel cycle take place in the Netherlands. These stages of the Dutch nuclear fuel cycle have been studied in detail and the external costs associated with thse stages estimated using up-to-date site specific data. The other stages of the Dutch nuclear fuel cycle do not currently take place in the Netherlands. In general the external costs associated with these stages have been estimated using data from the literature. Relatively few transports of radioactive materials associated with the Dutch nuclear fuel cycle take place in the Netherlands and the external costs associated with all transports has been based on values in the literature. (orig.).

  5. Modernization of the graphics post-processors of the Hamburg German Climate Computer Center Carbon Cycle Codes

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, E.J.; McNeilly, G.S.

    1994-03-01

    The existing National Center for Atmospheric Research (NCAR) code in the Hamburg Oceanic Carbon Cycle Circulation Model and the Hamburg Large-Scale Geostrophic Ocean General Circulation Model was modernized and reduced in size while still producing an equivalent end result. A reduction in the size of the existing code from more than 50,000 lines to approximately 7,500 lines in the new code has made the new code much easier to maintain. The existing code in Hamburg model uses legacy NCAR (including even emulated CALCOMP subrountines) graphics to display graphical output. The new code uses only current (version 3.1) NCAR subrountines.

  6. Modernization of the graphics post-processors of the Hamburg German Climate Computer Center Carbon Cycle Codes

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, E.J.; McNeilly, G.S.

    1994-03-01

    The existing National Center for Atmospheric Research (NCAR) code in the Hamburg Oceanic Carbon Cycle Circulation Model and the Hamburg Large-Scale Geostrophic Ocean General Circulation Model was modernized and reduced in size while still producing an equivalent end result. A reduction in the size of the existing code from more than 50,000 lines to approximately 7,500 lines in the new code has made the new code much easier to maintain. The existing code in Hamburg model uses legacy NCAR (including even emulated CALCOMP subrountines) graphics to display graphical output. The new code uses only current (version 3.1) NCAR subrountines.

  7. Dual Pressure versus Hybrid Recuperation in an Integrated Solid Oxide Fuel Cell Cycle – Steam Cycle

    DEFF Research Database (Denmark)

    Rokni, Masoud

    2014-01-01

    steam in a HRSG (heat recovery steam generator). The bottoming steam cycle was modeled with two configurations: (1) a simple single pressure level and (2) a dual pressure level with both a reheat and a pre-heater. The SOFC stacks in the present SOFC-ST hybrid cycles were not pressurized. The dual...

  8. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

    2016-11-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  9. Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

    Science.gov (United States)

    Rouf; Su'ud, Zaki

    2016-08-01

    Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

  10. 78 FR 71532 - Amendments to Material Control and Accounting Regulations and Proposed Guidance for Fuel Cycle...

    Science.gov (United States)

    2013-11-29

    ...: I. Background On November 8, 2013 (78 FR 67225; NRC-2009-0096), the NRC published for public comment a proposed rule to amend its regulations for MC&A of SNM. Also on November 8, 2013 (78 FR 67224; NRC... Control and Accounting Regulations and Proposed Guidance for Fuel Cycle Facility Material Control...

  11. Fuel Application Efficiency in Ideal Cycle of Gas Turbine Plant with Isobaric Heat Supply

    Directory of Open Access Journals (Sweden)

    A. Nesenchuk

    2013-01-01

    Full Text Available The paper reveals expediency to use in prospect fuels with maximum value  Qнр∑Vi and minimum theoretical burning temperature in order to obtain maximum efficiency of the ideal cycle in GTP with isobaric heat supply.

  12. 78 FR 11903 - Acceptability of Corrective Action Programs for Fuel Cycle Facilities

    Science.gov (United States)

    2013-02-20

    ... Commission. ACTION: Draft NUREG; request for public comment. SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is issuing for public comment draft NUREG-2154 ``Acceptability of Corrective Action Programs for Fuel Cycle Facilities.'' The draft NUREG provides guidance to the NRC staff on how to determine...

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

    NARCIS (Netherlands)

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

    2009-01-01

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

  14. 78 FR 23312 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

    Science.gov (United States)

    2013-04-18

    ... COMMISSION Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National... Energy Services (LES), LLC, National Enrichment Facility in Eunice, New Mexico, and has authorized the... Energy Act of 1954, as amended. The introduction of UF 6 into any module of the National...

  15. 78 FR 63518 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

    Science.gov (United States)

    2013-10-24

    ... COMMISSION Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National... conducted inspections of the Louisiana Energy Services (LES), LLC, National Enrichment Facility in Eunice... the Atomic Energy Act of 1954, as amended. The introduction of UF 6 into any module of the...

  16. Analysis of Trace Hydrogen Isotopes in Fusion Fuel Cycle by High Precision Gas Chromatograph

    Institute of Scientific and Technical Information of China (English)

    YANG; Li-ling; YANG; Hong-guang; ZHAO; Wei-wei; HE; Chang-shui; LIU; Zhen-xing; ZHAN; Qin

    2013-01-01

    It is essential to analysis of hydrogen isotopes in the fuel cycle system of fusion reactors,gas chromatography(GC)was found to be an effectively analytical technique.Compared with conventional GC,the high-precision GC with a cryogenic column could achieve fairly good performance to reduce the retention time.

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

    NARCIS (Netherlands)

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

    2009-01-01

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

  18. Back-end of the fuel cycle and non-proliferation strategies

    Energy Technology Data Exchange (ETDEWEB)

    Chebeskov, A.N.; Oussanov, V.I.; Iougai, S.V.; Pshakin, G.M. [Institute of Physics and Power Engineering, State Scientific Center of Russian Federation, Obninsk (Russian Federation)

    2001-07-01

    The paper focuses on the problem of fissile materials proliferation risk estimation. Some methodological approaches to the solution of this task and results of their application for comparison of different nuclear fuel cycle strategies are discussed. The results of comparative assessment of non-proliferation aspects of plutonium utilization alternatives in Russia using system analysis approach are presented. (author)

  19. To recycle or not to recycle? An intergenerational approach to nuclear fuel cycles.

    Science.gov (United States)

    Taebi, Behnam; Kloosterman, Jan Leen

    2008-06-01

    This paper approaches the choice between the open and closed nuclear fuel cycles as a matter of intergenerational justice, by revealing the value conflicts in the production of nuclear energy. The closed fuel cycle improve sustainability in terms of the supply certainty of uranium and involves less long-term radiological risks and proliferation concerns. However, it compromises short-term public health and safety and security, due to the separation of plutonium. The trade-offs in nuclear energy are reducible to a chief trade-off between the present and the future. To what extent should we take care of our produced nuclear waste and to what extent should we accept additional risks to the present generation, in order to diminish the exposure of future generation to those risks? The advocates of the open fuel cycle should explain why they are willing to transfer all the risks for a very long period of time (200,000 years) to future generations. In addition, supporters of the closed fuel cycle should underpin their acceptance of additional risks to the present generation and make the actual reduction of risk to the future plausible.

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

    NARCIS (Netherlands)

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

    2009-01-01

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

  1. Thermodynamic Analysis of an Integrated Gasification Solid Oxide Fuel Cell Plant with a Kalina Cycle

    DEFF Research Database (Denmark)

    Pierobon, Leonardo; Rokni, Masoud

    2015-01-01

    A hybrid plant that consists of a gasification system, Solid Oxide Fuel Cells (SOFC) and a Simple Kalina Cycle (SKC) is investigated. Woodchips are introduced into a fixed bed gasification plant to produce syngas, which is then fed into an integrated SOFC-SKC plant to produce electricity. The pre......-treated fuel then enters the anode side of the SOFC. Complete fuel oxidation is ensured in a burner by off-gases exiting the SOFC stacks. Off-gases are utilized as heat source for a SKC where a mixture of ammonia and water is expanded in a turbine to produce additional electric power. Thus, a triple novel...... system based on a gasification plant, a SOFC plant and a SKC plant is presented and investigated. The system is called IGSKC (Integrated Gasification SOFC Simple Kalina Cycle). The system layout is studied, and the optimal ammonia-water mole fraction is selected. An electrical efficiency of 58...

  2. Decision-making of biomass ethanol fuel policy based on life cycle 3E assessment

    Institute of Scientific and Technical Information of China (English)

    LENG Ru-bo; DAI Du; CHEN Xiao-jun; WANG Cheng-tao

    2005-01-01

    To evaluate the environmental, economic, energy performance of biomass ethanol fuel in China and tosupport the decision-making of biomass ethanol energy policy, an assessment method of life cycle 3E (economy, en vironment, energy) was applied to the three biomass ethanol fuel cycle alternatives, which includes cassava-based, corn-based and wheat-based ethanol fuel. The assessments provide a comparison of the economical performance, energy efficiency and environmental impacts of the three alternatives. And the development potential of the three alternatives in China was examined. The results are very useful for the Chinese government to make decisions on the biomass ethanol energy policy, and some advises for the decision-making of Chinese government were given.

  3. Measures of the Environmental Footprint of the Front End of the Nuclear Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Brett Carlsen; Emily Tavrides; Erich Schneider

    2010-08-01

    Previous estimates of environmental impacts associated with the front end of the nuclear fuel cycle have focused primarily on energy consumption and CO2 emissions. Results have varied widely. Section 2 of this report provides a summary of historical estimates. This study revises existing empirical correlations and their underlying assumptions to fit to a more complete set of existing data. This study also addresses land transformation, water withdrawals, and occupational and public health impacts associated with the processes of the front end of the once-through nuclear fuel cycle. These processes include uranium mining, milling, refining, conversion, enrichment, and fuel fabrication. Metrics are developed to allow environmental impacts to be summed across the full set of front end processes, including transportation and disposition of the resulting depleted uranium.

  4. Life-cycle analysis of bio-based aviation fuels.

    Science.gov (United States)

    Han, Jeongwoo; Elgowainy, Amgad; Cai, Hao; Wang, Michael Q

    2013-12-01

    Well-to-wake (WTWa) analysis of bio-based aviation fuels, including hydroprocessed renewable jet (HRJ) from various oil seeds, Fischer-Tropsch jet (FTJ) from corn-stover and co-feeding of coal and corn-stover, and pyrolysis jet from corn stover, is conducted and compared with petroleum jet. WTWa GHG emission reductions relative to petroleum jet can be 41-63% for HRJ, 68-76% for pyrolysis jet and 89% for FTJ from corn stover. The HRJ production stage dominates WTWa GHG emissions from HRJ pathways. The differences in GHG emissions from HRJ production stage among considered feedstocks are much smaller than those from fertilizer use and N2O emissions related to feedstock collection stage. Sensitivity analyses on FTJ production from coal and corn-stover are also conducted, showing the importance of biomass share in the feedstock, carbon capture and sequestration options, and overall efficiency. For both HRJ and FTJ, co-product handling methods have significant impacts on WTWa results. Copyright © 2013 Elsevier Ltd. All rights reserved.

  5. Comparison study of the thermal mechanical performance of fuel rods during BWR fuel preconditioning operations using the computer codes FUELSIM and FEMAXI-V

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Ingenieria Nuclear, Av. Instituto Politecnico Nacional s/n, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Ortiz V, J.; Castillo D, R., E-mail: rafael.pantoja10@yahoo.com.m [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico)

    2010-10-15

    The safety of nuclear power plants requires monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behaviour under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. In the operation of a nuclear power reactor, pre-conditioning simulations are necessary to determine in advance limit values for the power that can be generated in a fuel rod during any power ramp, and mainly during reactor startup, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR is performed. This study includes two types of fuel rods: one from a fuel assembly design with array 8 x 8, and the other one from a 10 x 10 fuel assembly design, and a comparison of the thermal-mechanical performance between the two different rod designs is performed. The performance simulations were performed by the code FUELSIM, and compared against results previously obtained from similar simulation with the code FEMAXI-V. (Author)

  6. Monte-Carlo code calculation of 3D reactor core model with usage of burnt fuel isotopic compositions, obtained by engineering codes

    Energy Technology Data Exchange (ETDEWEB)

    Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

    2016-09-15

    A burn-up calculation of large systems by Monte-Carlo code (MCU) is complex process and it requires large computational costs. Previously prepared isotopic compositions are proposed to be used for the Monte-Carlo code calculations of different system states with burnt fuel. Isotopic compositions are calculated by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by the engineering codes (TVS-M, BIPR-7A and PERMAK-A). The multiplication factors and power distributions of FAs from a 3-D reactor core are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The separate conditions of the burnt core are observed. The results of MCU calculations were compared with those that were obtained by engineering codes.

  7. Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Čerba, Štefan, E-mail: stefan.cerba@stuba.sk [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia); Vrban, Branislav; Lüley, Jakub [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia); Dařílek, Petr [VUJE a.s., Okružná 5, 918 64 Trnava (Slovakia); Zajac, Radoslav, E-mail: radoslav.zajac@vuje.sk [VUJE a.s., Okružná 5, 918 64 Trnava (Slovakia); Nečas, Vladimír; Haščik, Ján [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia)

    2014-02-15

    Highlights: • Verification of the MCNPX, HELIOS and SCALE codes. • MOX and ceramic fuel assembly. • Gas-cooled fast reactor. • Burnup calculation. - Abstract: The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice.

  8. Hydraulic Hybrid and Conventional Parcel Delivery Vehicles' Measured Laboratory Fuel Economy on Targeted Drive Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Lammert, M. P.; Burton, J.; Sindler, P.; Duran, A.

    2014-10-01

    This research project compares laboratory-measured fuel economy of a medium-duty diesel powered hydraulic hybrid vehicle drivetrain to both a conventional diesel drivetrain and a conventional gasoline drivetrain in a typical commercial parcel delivery application. Vehicles in this study included a model year 2012 Freightliner P100H hybrid compared to a 2012 conventional gasoline P100 and a 2012 conventional diesel parcel delivery van of similar specifications. Drive cycle analysis of 484 days of hybrid parcel delivery van commercial operation from multiple vehicles was used to select three standard laboratory drive cycles as well as to create a custom representative cycle. These four cycles encompass and bracket the range of real world in-use data observed in Baltimore United Parcel Service operations. The NY Composite cycle, the City Suburban Heavy Vehicle Cycle cycle, and the California Air Resources Board Heavy Heavy-Duty Diesel Truck (HHDDT) cycle as well as a custom Baltimore parcel delivery cycle were tested at the National Renewable Energy Laboratory's Renewable Fuels and Lubricants Laboratory. Fuel consumption was measured and analyzed for all three vehicles. Vehicle laboratory results are compared on the basis of fuel economy. The hydraulic hybrid parcel delivery van demonstrated 19%-52% better fuel economy than the conventional diesel parcel delivery van and 30%-56% better fuel economy than the conventional gasoline parcel delivery van on cycles other than the highway-oriented HHDDT cycle.

  9. Uncertainty in life cycle economical analysis of cassava-based ethanol fuel

    Institute of Scientific and Technical Information of China (English)

    LENG Ru-bo; DAI Du; CHEN Xiao-jun; WANG Cheng-tao

    2005-01-01

    Biomass ethanol fuel is not only renewable but also environmental-friendly. Guangxi Zhuang Autonomous Region is developing the cassava-based ethanol fuel. Economical performance of the project is the key issue.The traditional life cycle economical analysis is just a static calculation process. Uncertainty is the character of cassava yield, cost of cassava plant, cassava price, tax rate and gasoline price, and the economical performance of the project is determined by these aspects. This study proposes an economical model of cassava-based ethanol fuel. The method of Monte Carol is used to simulate the economical performance. This method conquers the shortage of the traditional way. The results show that cassava-based ethanol fuel can get survived when the tax is exempted. Finally, the study also evaluates the potential of the economical performance.

  10. SOLVENT EXTRACTION RESEARCH AND DEVELOPMENT IN THE U.S. FUEL CYCLE PROGRAM

    Energy Technology Data Exchange (ETDEWEB)

    Terry A. Todd

    2011-10-01

    Treatment or processing of used nuclear fuel to recycle uranium and plutonium has historically been accomplished using the well known PUREX process. The PUREX process has been used on an industrial scale for over 60 years in the nuclear industry. Research is underway to develop advanced separation methods for the recovery of other used fuel components, such as the minor actinides (Np, Am, Cm) for possible transmutation in fast spectrum reactors, or other constituents (e.g. Cs, Sr, transition metals, lanthanides) to help facilitate effective waste management options. This paper will provide an overview of new solvent extraction processes developed for advanced nuclear fuel cycles, and summarize recent experimental results. This will include the utilization of new extractants for selective separation of target metals and new processes developed to selectively recover one or more elements from used fuel.

  11. Molten salt considerations for accelerator-driven subcritical fission to close the nuclear fuel cycle

    Science.gov (United States)

    Sooby, Elizabeth; Adams, Marvin; Baty, Austin; Gerity, James; McIntyre, Peter; Melconian, Karie; Phongikaroon, Supathorn; Pogue, Nathaniel; Sattarov, Akhdiyor; Simpson, Michael; Tripathy, Prabhat; Tsevkov, Pavel

    2013-04-01

    The host salt selection, molecular modeling, physical chemistry, and processing chemistry are presented here for an accelerator-driven subcritical fission in a molten salt core (ADSMS). The core is fueled solely with the transuranics (TRU) and long-lived fission products (LFP) from used nuclear fuel. The neutronics and salt composition are optimized to destroy the transuranics by fission and the long-lived fission products by transmutation. The cores are driven by proton beams from a strong-focusing cyclotron stack. One such ADSMS system can destroy the transuranics in the used nuclear fuel produced by a 1GWe conventional reactor. It uniquely provides a method to close the nuclear fuel cycle for green nuclear energy.

  12. Life cycle assessment of camelina oil derived biodiesel and jet fuel in the Canadian Prairies.

    Science.gov (United States)

    Li, Xue; Mupondwa, Edmund

    2014-05-15

    This study evaluated the environmental impact of biodiesel and hydroprocessed renewable jet fuel derived from camelina oil in terms of global warming potential, human health, ecosystem quality, and energy resource consumption. The life cycle inventory is based on production activities in the Canadian Prairies and encompasses activities ranging from agricultural production to oil extraction and fuel conversion. The system expansion method is used in this study to avoid allocation and to credit input energy to co-products associated with the products displaced in the market during camelina oil extraction and fuel processing. This is the preferred allocation method for LCA analysis in the context of most renewable and sustainable energy programs. The results show that greenhouse gas (GHG) emissions from 1 MJ of camelina derived biodiesel ranged from 7.61 to 24.72 g CO2 equivalent and 3.06 to 31.01 kg CO2/MJ equivalent for camelina HRJ fuel. Non-renewable energy consumption for camelina biodiesel ranged from 0.40 to 0.67 MJ/MJ; HRJ fuel ranged from -0.13 to 0.52 MJ/MJ. Camelina oil as a feedstock for fuel production accounted for the highest contribution to overall environmental performance, demonstrating the importance of reducing environmental burdens during the agricultural production process. Attaining higher seed yield would dramatically lower environmental impacts associated with camelina seed, oil, and fuel production. The lower GHG emissions and energy consumption associated with camelina in comparison with other oilseed derived fuel and petroleum fuel make camelina derived fuel from Canadian Prairies environmentally attractive.

  13. Fuel Cycle Research & Development Technical Monthly - June 2012

    Energy Technology Data Exchange (ETDEWEB)

    Miller, Michael C. [Los Alamos National Laboratory

    2012-07-25

    two papers to the upcoming INMM meeting that are related to the fast neutron multiplicity R&D effort. Progress was made on the project's main goal of designing a concept for a prototype fast-neutron multiplicity counter. We started laying out the outline for the final report. We have been working with our ORNL collaborators to develop a new digitizer system to support our experimental campaign planned for next year. (3) MPACT Analysis Tools - (a) Multi-isotope Process Monitor - Fuel characterization framework development continued during June. A report describing the methodologies is being completed. Kenneth Dayman, from University of Texas, spent a week at PNNL wrapping up his master's research and working on a journal submission covering that work. The target journal is the IEEE transactions on Nuclear Science; submission is planned for the end of July. A proposal to instrument H-Canyon is being prepared in conjunction with SRNL and the NNSA's NGSI program. The impact of gamma-ray spectrum counting statistics on the precision of relative radioisotope component intensities as reconstructed via Principal Component Regression (PCR) continued in June with Monte Carlo simulations of a two-component (i.e., two radioisotope) system. This work generalizes earlier studies in FY12 in which Poisson counting variations of only a single spectrum component were simulated. (b) Modeling and Simulation for Analysis of Safeguards Performance (Electrochemical) - Preliminary insights into safeguards challenges and the initial design for an electrochemical plant have been written up into an INMM paper and will be presented at the INMM Summer Meeting. Work is currently adding a new visualization capability for integrating materials accountancy with physical protection. (c) Material Control including Process Monitoring (Pattern Recognition, Sensors) - Fabrication of quartz chips continued at an external foundry. Awaiting delivery of the heat exchange manifold and chip

  14. Model improvements for tritium transport in DEMO fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Santucci, Alessia, E-mail: alessia.santucci@enea.it [Unità Tecnica Fusione – ENEA C. R. Frascati, Via E. Fermi 45, 00044 Frascati (Roma) (Italy); Tosti, Silvano [Unità Tecnica Fusione – ENEA C. R. Frascati, Via E. Fermi 45, 00044 Frascati (Roma) (Italy); Franza, Fabrizio [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

    2015-10-15

    Highlights: • T inventory and permeation of DEMO blankets have been assessed under pulsed operation. • 1-D model for T transport has been developed for the HCLL DEMO blanket. • The 1-D model evaluated T partial pressure and T permeation rate radial profiles. - Abstract: DEMO operation requires a large amount of tritium, which is directly produced inside the reactor by means of Li-based breeders. During its production, recovering and purification, tritium comes in contact with large surfaces of hot metallic walls, therefore it can permeate through the blanket cooling structure, reach the steam generator and finally the environment. The development of dedicated simulation tools able to predict tritium losses and inventories is necessary to verify the accomplishment of the accepted tritium environmental releases as well as to guarantee a correct machine operation. In this work, the FUS-TPC code is improved by including the possibility to operate in pulsed regime: results in terms of tritium inventory and losses for three pulsed scenarios are shown. Moreover, the development of a 1-D model considering the radial profile of the tritium generation is described. By referring to the inboard segment on the equatorial axis of the helium-cooled lithium–lead (HCLL) blanket, preliminary results of the 1-D model are illustrated: tritium partial pressure in Li–Pb and tritium permeation in the cooling and stiffening plates by assuming several permeation reduction factor (PRF) values. Future improvements will consider the application of the model to all segments of different blanket concepts.

  15. Life-cycle cost analysis of conventional and fuel cell/battery powered urban passenger vehicles

    Science.gov (United States)

    1992-11-01

    This Final Report summarizes the work on the life cycle cost (LCC) analysis of conventional and fuel cell/battery powered urban passenger vehicles. The purpose of the work is to support the Division in making sound economic comparisons between conventional and fuel cell/battery powered buses, passenger vans, and cars for strategic analysis of programmatic R&D goals. The LCC analysis can indicate whether paying a relatively high initial capital cost for advanced technology with low operating and/or environmental costs is advantageous over paying a lower initial cost for conventional technology with higher operating and/or environmental costs. While minimizing life cycle cost is an important consideration, it does not always result in technology penetration in the marketplace. The LCC analysis model developed under this contract facilitates consideration of all perspectives. Over 100 studies have been acquired and analyzed for their applicability. Drawing on prior work by JPL and Los Alamos National Laboratory as primary sources, specific analytical relationships and cost/performance data relevant to fuel cell/battery and intemal combustion engine (ICE) powered vehicles were selected for development of an LCC analysis model. The completed LCC model is structured around twelve integrated modules. Comparative analysis is made between conventional gasoline and diesel vehicles and fuel cell/battery vehicles using either phosphoric acid fuel cells or proton-exchange membrane fuel cells. In all, seven base vehicle configuration cases with a total of 21 vehicle class/powertrain/fuel combinations are analyzed. The LCC model represents a significant advance in comparative economic analysis of conventional and fuel cell/battery powered vehicle technologies embodying several unique features which were not included in prior models.

  16. Evaluation of fretting failures on PWR fuel by post-irradiation examinations and modeling in the DEGRAD-1 code

    Energy Technology Data Exchange (ETDEWEB)

    Castanheira, Myrthes; Silva, Jose Eduardo Rosa da; Lucki, Georgi; Terremoto, Luis A.A.; Silva, Antonio Teixeira e; Teodoro, Celso A.; Damy, Margaret de A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mail: myrthes@ipen.br

    2007-07-01

    One of the major recognized causes of fuel rod failures is fretting of the clad due to the entrapment of debris in a fuel rod spacer. Such debris, inadvertently dropped into the primary system during maintenance operations, includes various sizes of particles. Intermediate size particles, such as metal cuttings, electrical connectors, metal fittings, pieces of wire, and small nuts and bolts can become trapped between fuel rods in a spacer where hydraulically induced vibrations can cause fretting failure of the fuel rod. An evaluation of debris fretting failure on PWR fuel is presented. The inquiries on fuel rods failures are based on results of analysis using post-irradiation non-destructive examination. The complementary analysis includes a modeling approach by code DEGRAD-1 to characterize the degradation phenomenon after primary failure integrated in the reactor operational history. (author)

  17. Work plan for improving the DARWIN2.3 depleted material balance calculation of nuclides of interest for the fuel cycle

    Science.gov (United States)

    Rizzo, Axel; Vaglio-Gaudard, Claire; Martin, Julie-Fiona; Noguère, Gilles; Eschbach, Romain

    2017-09-01

    DARWIN2.3 is the reference package used for fuel cycle applications in France. It solves the Boltzmann and Bateman equations in a coupling way, with the European JEFF-3.1.1 nuclear data library, to compute the fuel cycle values of interest. It includes both deterministic transport codes APOLLO2 (for light water reactors) and ERANOS2 (for fast reactors), and the DARWIN/PEPIN2 depletion code, each of them being developed by CEA/DEN with the support of its industrial partners. The DARWIN2.3 package has been experimentally validated for pressurized and boiling water reactors, as well as for sodium fast reactors; this experimental validation relies on the analysis of post-irradiation experiments (PIE). The DARWIN2.3 experimental validation work points out some isotopes for which the depleted concentration calculation can be improved. Some other nuclides have no available experimental validation, and their concentration calculation uncertainty is provided by the propagation of a priori nuclear data uncertainties. This paper describes the work plan of studies initiated this year to improve the accuracy of the DARWIN2.3 depleted material balance calculation concerning some nuclides of interest for the fuel cycle.

  18. MUICYCL and MUIFAP: models tracking minor uranium isotopes in the nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Blum, S.R.; McLaren, R.A.

    1979-10-01

    Two computer programs have been written to provide information on the buildup of minor uranium isotopes in the nuclear fuel cycle. The Minor Uranium Isotope Cycle Program, MUICYCL, tracks fuel through a multiyear campaign cycle of enrichment, reactor burnup, reprocessing, enrichment, etc. MUICYCL facilities include preproduction stockpiles, U/sup 235/ escalation, and calculation of losses. The Minor Uranium Isotope Flowsheet Analyzer Program, MUIFAP, analyzes one minor isotope in one year of an enrichment operation. The formulation of the enrichment cascade, reactors, and reprocessing facility is presented. Input and output descriptions and sample cases are presented. The programs themselves are documented by short descriptions of each routine, flowcharts, definitions of common blocks and variables, and internal documentation. The programs are written in FORTRAN for use in batch mode.

  19. A Mathematical Model for Predicting the Life of PEM Fuel Cell Membranes Subjected to Hydration Cycling

    CERN Document Server

    Burlatsky, S F; O'Neill, J; Atrazhev, V V; Varyukhin, A N; Dmitriev, D V; Erikhman, N S

    2013-01-01

    Under typical PEM fuel cell operating conditions, part of membrane electrode assembly is subjected to humidity cycling due to variation of inlet gas RH and/or flow rate. Cyclic membrane hydration/dehydration would cause cyclic swelling/shrinking of the unconstrained membrane. In a constrained membrane, it causes cyclic stress resulting in mechanical failure in the area adjacent to the gas inlet. A mathematical modeling framework for prediction of the lifetime of a PEM FC membrane subjected to hydration cycling is developed in this paper. The model predicts membrane lifetime as a function of RH cycling amplitude and membrane mechanical properties. The modeling framework consists of three model components: a fuel cell RH distribution model, a hydration/dehydration induced stress model that predicts stress distribution in the membrane, and a damage accrual model that predicts membrane life-time. Short descriptions of the model components along with overall framework are presented in the paper. The model was used...

  20. Formulating energy policies related to fossil fuel use: Critical uncertainties in the global carbon cycle

    Energy Technology Data Exchange (ETDEWEB)

    Post, W.M.; Dale, V.H.; DeAngelis, D.L.; Mann, L.K.; Mulholland, P.J.; O' Neill, R.V.; Peng, T.-H.; Farrell, M.P.

    1990-01-01

    The global carbon cycle is the dynamic interaction among the earth's carbon sources and sinks. Four reservoirs can be identified, including the atmosphere, terrestrial biosphere, oceans, and sediments. Atmospheric CO{sub 2} concentration is determined by characteristics of carbon fluxes among major reservoirs of the global carbon cycle. The objective of this paper is to document the knowns, and unknowns and uncertainties associated with key questions that if answered will increase the understanding of the portion of past, present, and future atmospheric CO{sub 2} attributable to fossil fuel burning. Documented atmospheric increases in CO{sub 2} levels are thought to result primarily from fossil fuel use and, perhaps, deforestation. However, the observed atmospheric CO{sub 2} increase is less than expected from current understanding of the global carbon cycle because of poorly understood interactions among the major carbon reservoirs. 87 refs.

  1. Formulating Energy Policies Related to Fossil Fuel Use: Critical Uncertainties in the Global Carbon Cycle

    Science.gov (United States)

    Post, W. M.; Dale, V. H.; DeAngelis, D. L.; Mann, L. K.; Mulholland, P. J.; O`Neill, R. V.; Peng, T. -H.; Farrell, M. P.

    1990-02-01

    The global carbon cycle is the dynamic interaction among the earth's carbon sources and sinks. Four reservoirs can be identified, including the atmosphere, terrestrial biosphere, oceans, and sediments. Atmospheric CO{sub 2} concentration is determined by characteristics of carbon fluxes among major reservoirs of the global carbon cycle. The objective of this paper is to document the knowns, and unknowns and uncertainties associated with key questions that if answered will increase the understanding of the portion of past, present, and future atmospheric CO{sub 2} attributable to fossil fuel burning. Documented atmospheric increases in CO{sub 2} levels are thought to result primarily from fossil fuel use and, perhaps, deforestation. However, the observed atmospheric CO{sub 2} increase is less than expected from current understanding of the global carbon cycle because of poorly understood interactions among the major carbon reservoirs.

  2. Combinations of solid oxide fuel cell and several enhanced gas turbine cycles

    Science.gov (United States)

    Kuchonthara, Prapan; Bhattacharya, Sankar; Tsutsumi, Atsushi

    Combined power generation systems with combinations of solid oxide fuel cell (SOFC) and various enhanced gas turbine (GT) cycles were evaluated. In the GT part, steam injected gas turbine (STIG) cycle, GT/steam turbine (ST) combined cycle, and humid air turbine (HAT) cycle were considered. Moreover, additional recuperation was considered by means of air preheating (APH) in the STIG cycle. Effects of operating turbine inlet temperature (TIT) and pressure ratio (PR) on overall system performance were assessed. Although the SOFC-HAT system shows the lowest specific work output compared to other systems, its highest thermal efficiency presents a significant advantage. Furthermore, at high TITs and PRs the SOFC-HAT system gives the best performance in terms of both thermal efficiency and specific work. Results indicate that energy recuperative features in the HAT promote the positive effect of increasing TIT by means of enhancing GT efficiency, leading to the improvement in thermal efficiency of the overall system.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  4. A step towards closing the CANDU fuel cycle: an innovative scheme for reprocessing used CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Collins, F.; Lister, D. [Univ. of New Brunswick, UNB Nuclear, Dept. of Chemical Engineering, Fredericton, New Brunswick (Canada)

    2011-07-01

    Disposal versus reprocessing costs for used CANDU fuel was recently discussed by Rozon and Lister in a report produced for the Nuclear Waste Management Organization (NWMO). Their study discussed the economic incentives for reprocessing, not for the recovery of fissile uranium but for the recovery of plutonium ash. A $370/kg break-even price of uranium was calculated, and their model was found to be very sensitive to the reprocessing costs of the chosen technology. Findings were consistent with earlier studies done by Harvard University. Various reprocessing technologies (most based on solvent extraction) have been in use for many decades, but there appears to be no conceptual engineering study available in the open literature for a spent fuel reprocessing facility - one that includes process flows, operating costs and economic analysis. A deeper engineering study of the design and economics of re-processing technologies has since been undertaken by the nuclear group at the University of New Brunswick. An improved fluorination process was developed and modeled using ASPEN process simulation software. This study examines the impact of chosen technology on the spent fuel re-processing costs. (author)

  5. Cycling experiences and knowledge of the road code of nine-year-olds.

    Science.gov (United States)

    Langley, J; Silva, P A; Williams, S M

    1987-04-01

    Sex differences in cycling experience and knowledge of the road code were examined in a sample of eight-hundred and fifteen nine-year-old children. The results showed that there were not significant sex differences for those who had ridden on the road or ridden to school in the week preceding the interview. However, males were more likely to have ridden at night. There were no significant sex differences in self reported unsafe acts or in knowledge of various dimensions of the road code. The results show a relatively high level of ignorance of the give-way rules for uncontrolled intersections and a disturbing number of children who reported they had ridden at night without tail or headlights.

  6. Analysis of Pu-Only Partitioning Strategies in LMFBR Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Samuel Bays; Gilles Youinou

    2013-02-01

    Sodium cooled Fast Reactors (SFR) have been under consideration for production of electricity, fissile material production, and for destruction of transuranics for decades. The neutron economy of a SFR can be operated in one of two ways. One possibility is to operate the reactor in a transuranic burner mode which has been the focus of active R&D in the last 15 years. However, prior to that the focus was on breeding transuranics. This later mode of managing the neutron economy relies on ensuring the maximum fuel utilization possible in such a way as to maximize the amount of plutonium produced per unit of fission energy in the reactor core. The goal of maximizing plutonium production in this study is as fissile feed stock for the production of MOX fuel to be used in Light Water Reactors (LWR). Throughout the l970’s, this fuel cycle scenario was the focus of much research by the Atomic Energy Commission in the event that uranium supplies would be scarce. To date, there has been sufficient uranium to supply the once through nuclear fuel cycle. However, interest in a synergistic relationship Liquid Metal Fast Breeder Reactors (LMFBR) and a consumer LWR fleet persists, prompting this study. This study considered LMFBR concepts with varying additions of axial and radial reflectors. Three scenarios were considered in collaboration with a companion study on the LWR-MOX designs based on plutonium nuclide vectors produced by this study. The first scenario is a LMFBR providing fissile material to make MOX fuel where the MOX part of the fuel cycle is operated in a once-through-then-out mode. The second scenario is the same as the first but with the MOX part of the fuel cycle multi-recycling its own plutonium with LMFBR being used for the make-up feed. In these first two scenarios, plutonium partitioning from the minor actinides (MA) was assumed. Also, the plutonium management strategy of the LMFBR ensured that only the high fissile purity plutonium bred from blankets was

  7. Compressor Modeling for Transient Analysis of Supercritical CO2 Brayton Cycle by using MARS code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hyun; Park, Hyun Sun; Kim, Tae Ho; Kwon, Jin Gyu [POSTECH, Pohang (Korea, Republic of); Bae, Sung Won; Cha, Jae Eun [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, SCIEL (Supercritical CO{sub 2} Integral Experimental Loop) was chosen as a reference loop and the MARS code was as the transient cycle analysis code. As a result, the compressor homologous curve was developed from the SCIEL experimental data and MARS analysis was performed and presented in the paper. The advantages attract SCO{sub 2}BC as a promising next generation power cycles. The high thermal efficiency comes from the operation of compressor near the critical point where the properties of SCO{sub 2}. The approaches to those of liquid phase, leading drastically lower the compression work loss. However, the advantage requires precise and smooth operation of the cycle near the critical point. However, it is one of the key technical challenges. The experimental data was steady state at compressor rotating speed of 25,000 rpm. The time, 3133 second, was starting point of steady state. Numerical solutions were well matched with the experimental data. The mass flow rate from the MARS analysis of approximately 0.7 kg/s was close to the experimental result of 0.9 kg/s. It is expected that the difference come from the measurement error in the experiment. In this study, the compressor model was developed and implemented in MARS to study the transient analysis of SCO{sub 2}BC in SCIEL. We obtained the homologous curves for the SCIEL compressor using experimental data and performed nodalization of the compressor model using MARS code. In conclusions, it was found that numerical solutions from the MARS model were well matched with experimental data.

  8. Nuclear fuel cycle risk assessment: survey and computer compilation of risk-related literature. [Once-through Cycle and Plutonium Recycle

    Energy Technology Data Exchange (ETDEWEB)

    Yates, K.R.; Schreiber, A.M.; Rudolph, A.W.

    1982-10-01

    The US Nuclear Regulatory Commission has initiated the Fuel Cycle Risk Assessment Program to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. Both the once-through cycle and plutonium recycle are being considered. A previous report generated by this program defines and describes fuel cycle facilities, or elements, considered in the program. This report, the second from the program, describes the survey and computer compilation of fuel cycle risk-related literature. Sources of available information on the design, safety, and risk associated with the defined set of fuel cycle elements were searched and documents obtained were catalogued and characterized with respect to fuel cycle elements and specific risk/safety information. Both US and foreign surveys were conducted. Battelle's computer-based BASIS information management system was used to facilitate the establishment of the literature compilation. A complete listing of the literature compilation and several useful indexes are included. Future updates of the literature compilation will be published periodically. 760 annotated citations are included.

  9. Study of Thorium-Plutonium Fuel for Possible Operating Cycle Extension in PWRs

    Directory of Open Access Journals (Sweden)

    Klara Insulander Björk

    2013-01-01

    Full Text Available Computer simulations have been carried out to investigate the possibility of extending operating cycle length in the Pressurised Water Reactor Ringhals 3 by the use of thorium-plutonium oxide fuel. The calculations have been carried out using tools and methods that are normally employed for reload design and safety evaluation in Ringhals 3. The 3-batch reload scheme and the power level have been kept unchanged, and a normal uranium oxide fuel assembly designed for a 12-month operating cycle in this reactor is used as a reference. The use of plutonium as the fissile component reduces the worth of control rods and soluble boron, which makes it necessary to modify the control systems. The delayed neutron fraction is low compared with the reference, but simulations and qualitative assessments of relevant transients indicate that the reactor could still be operate