Radiation Monitoring System in Advanced Spent Fuel Conditioning Process Facility

International Nuclear Information System (INIS)

You, Gil Sung; Kook, D. H.; Choung, W. M.; Ku, J. H.; Cho, I. J.; You, G. S.; Kwon, K. C.; Lee, W. K.; Lee, E. P.

2006-09-01

The Advanced spent fuel Conditioning Process is under development for effective management of spent fuel by converting UO 2 into U-metal. For demonstration of this process, α-γ type new hot cell was built in the IMEF basement . To secure against radiation hazard, this facility needs radiation monitoring system which will observe the entire operating area before the hot cell and service area at back of it. This system consists of 7 parts; Area Monitor for γ-ray, Room Air Monitor for particulate and iodine in both area, Hot cell Monitor for hot cell inside high radiation and rear door interlock, Duct Monitor for particulate of outlet ventilation, Iodine Monitor for iodine of outlet duct, CCTV for watching workers and material movement, Server for management of whole monitoring system. After installation and test of this, radiation monitoring system will be expected to assist the successful ACP demonstration

The miscibility and oxidation study of the simulated metallic spent fuel for the development of an advanced spent fuel management process

Energy Technology Data Exchange (ETDEWEB)

Shin, Y. J.; You, G. S.; Ju, J. S.; Lee, E. P.; Seo, H. S.; Ahn, S. B. [Korea Atomic Energy Research Institute, Taejeon (Korea)

1999-03-01

The simulated metallic spent fuel ingots were fabricated and evaluated the oxidation rates and the activation energies under several temperature conditions to develop an advanced spent fuel management process. It was also checked the immiscibility of the some elements with metal uranium. 2 refs., 45 figs. (Author)

Advanced Fuels Campaign FY 2011 Accomplishments Report

Energy Technology Data Exchange (ETDEWEB)

Not Listed

2011-11-01

One of the major research and development (R&D) areas under the Fuel Cycle Research and Development (FCRD) program is advanced fuels development. The Advanced Fuels Campaign (AFC) has the responsibility to develop advanced fuel technologies for the Department of Energy (DOE) using a science-based approach focusing on developing a microstructural understanding of nuclear fuels and materials. Accomplishments made during fiscal year (FY 20) 2011 are highlighted in this report, which focuses on completed work and results. The process details leading up to the results are not included; however, the technical contact is provided for each section. The order of the accomplishments in this report is consistent with the AFC work breakdown structure (WBS).

Development of Advanced Spent Fuel Management Process

Energy Technology Data Exchange (ETDEWEB)

Seo, Chung Seok; Choi, I. K.; Kwon, S. G. (and others)

2007-06-15

As a part of research efforts to develop an advanced spent fuel management process, this project focused on the electrochemical reduction technology which can replace the original Li reduction technology of ANL, and we have successfully built a 20 kgHM/batch scale demonstration system. The performance tests of the system in the ACPF hot cell showed more than a 99% reduction yield of SIMFUEL, a current density of 100 mA/cm{sup 2} and a current efficiency of 80%. For an optimization of the process, the prevention of a voltage drop in an integrated cathode, a minimization of the anodic effect and an improvement of the hot cell operability by a modulation and simplization of the unit apparatuses were achieved. Basic research using a bench-scale system was also carried out by focusing on a measurement of the electrochemical reduction rate of the surrogates, an elucidation of the reaction mechanism, collecting data on the partition coefficients of the major nuclides, quantitative measurement of mass transfer rates and diffusion coefficients of oxygen and metal ions in molten salts. When compared to the PYROX process of INL, the electrochemical reduction system developed in this project has comparative advantages in its application of a flexible reaction mechanism, relatively short reaction times and increased process yields.

Development of Advanced Spent Fuel Management Process

International Nuclear Information System (INIS)

Seo, Chung Seok; Choi, I. K.; Kwon, S. G.

2007-06-01

As a part of research efforts to develop an advanced spent fuel management process, this project focused on the electrochemical reduction technology which can replace the original Li reduction technology of ANL, and we have successfully built a 20 kgHM/batch scale demonstration system. The performance tests of the system in the ACPF hot cell showed more than a 99% reduction yield of SIMFUEL, a current density of 100 mA/cm 2 and a current efficiency of 80%. For an optimization of the process, the prevention of a voltage drop in an integrated cathode, a minimization of the anodic effect and an improvement of the hot cell operability by a modulation and simplization of the unit apparatuses were achieved. Basic research using a bench-scale system was also carried out by focusing on a measurement of the electrochemical reduction rate of the surrogates, an elucidation of the reaction mechanism, collecting data on the partition coefficients of the major nuclides, quantitative measurement of mass transfer rates and diffusion coefficients of oxygen and metal ions in molten salts. When compared to the PYROX process of INL, the electrochemical reduction system developed in this project has comparative advantages in its application of a flexible reaction mechanism, relatively short reaction times and increased process yields

Advanced fuels safety comparisons

International Nuclear Information System (INIS)

Grolmes, M.A.

1977-01-01

The safety considerations of advanced fuels are described relative to the present understanding of the safety of oxide fueled Liquid Metal Fast Breeder Reactors (LMFBR). Safety considerations important for the successful implementation of advanced fueled reactors must early on focus on the accident energetics issues of fuel coolant interactions and recriticality associated with core disruptive accidents. It is in these areas where the thermal physical property differences of the advanced fuel have the greatest significance

Advanced methods of process/quality control in nuclear reactor fuel manufacture. Proceedings of a technical committee meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-07-01

Nuclear fuel plays an essential role in ensuring the competitiveness of nuclear energy and its acceptance by the public. The economic and market situation is not favorable at present for nuclear fuel designers and suppliers. The reduction in fuel prices (mainly to compete with fossil fuels) and in the number of fuel assemblies to be delivered to customers (mainly due to burnup increase) has been offset by the rising number of safety and other requirements, e.g. the choice of fuel and structural materials and the qualification of equipment. In this respect, higher burnup and thermal rates, longer fuel cycles and the use of MOX fuels are the real means to improve the economics of the nuclear fuel cycle as a whole. Therefore, utilities and fuel vendors have recently initiated new research and development programmes aimed at improving fuel quality, design and materials to produce robust and reliable fuel for safe and reliable reactor operation more demanding conditions. In this connection, improvement of fuel quality occupies an important place and this requires continuous effort on the part of fuel researchers, designers and producers. In the early years of commercial fuel fabrication, emphasis was given to advancements in quality control/quality assurance related mainly to the product itself. Now, the emphasis is transferred to improvements in process control and to implementation of overall total quality management (TQM) programmes. In the area of fuel quality control, statistical methods are now widely implemented, replacing 100% inspection. The IAEA, recognizing the importance of obtaining and maintaining high standards in fuel fabrication, has paid particular attention to this subject. In response to the rapid progress in development and implementation of advanced methods of process/quality control in nuclear fuel manufacture and on the recommendation of the International Working Group on Water Reactor Fuel Performance and Technology, the IAEA conducted a

Advanced methods of process/quality control in nuclear reactor fuel manufacture. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

2000-07-01

Nuclear fuel plays an essential role in ensuring the competitiveness of nuclear energy and its acceptance by the public. The economic and market situation is not favorable at present for nuclear fuel designers and suppliers. The reduction in fuel prices (mainly to compete with fossil fuels) and in the number of fuel assemblies to be delivered to customers (mainly due to burnup increase) has been offset by the rising number of safety and other requirements, e.g. the choice of fuel and structural materials and the qualification of equipment. In this respect, higher burnup and thermal rates, longer fuel cycles and the use of MOX fuels are the real means to improve the economics of the nuclear fuel cycle as a whole. Therefore, utilities and fuel vendors have recently initiated new research and development programmes aimed at improving fuel quality, design and materials to produce robust and reliable fuel for safe and reliable reactor operation more demanding conditions. In this connection, improvement of fuel quality occupies an important place and this requires continuous effort on the part of fuel researchers, designers and producers. In the early years of commercial fuel fabrication, emphasis was given to advancements in quality control/quality assurance related mainly to the product itself. Now, the emphasis is transferred to improvements in process control and to implementation of overall total quality management (TQM) programmes. In the area of fuel quality control, statistical methods are now widely implemented, replacing 100% inspection. The IAEA, recognizing the importance of obtaining and maintaining high standards in fuel fabrication, has paid particular attention to this subject. In response to the rapid progress in development and implementation of advanced methods of process/quality control in nuclear fuel manufacture and on the recommendation of the International Working Group on Water Reactor Fuel Performance and Technology, the IAEA conducted a

The development of technical database of advanced spent fuel management process

Energy Technology Data Exchange (ETDEWEB)

Ro, Seung Gy; Byeon, Kee Hoh; Song, Dae Yong; Park, Seong Won; Shin, Young Jun

1999-03-01

The purpose of this study is to develop the technical database system to provide useful information to researchers who study on the back end nuclear fuel cycle. Technical database of advanced spent fuel management process was developed for a prototype system in 1997. In 1998, this database system is improved into multi-user systems and appended special database which is composed of thermochemical formation data and reaction data. In this report, the detailed specification of our system design is described and the operating methods are illustrated as a user's manual. Also, expanding current system, or interfacing between this system and other system, this report is very useful as a reference. (Author). 10 refs., 18 tabs., 46 fig.

The development of technical database of advanced spent fuel management process

International Nuclear Information System (INIS)

Ro, Seung Gy; Byeon, Kee Hoh; Song, Dae Yong; Park, Seong Won; Shin, Young Jun

1999-03-01

The purpose of this study is to develop the technical database system to provide useful information to researchers who study on the back end nuclear fuel cycle. Technical database of advanced spent fuel management process was developed for a prototype system in 1997. In 1998, this database system is improved into multi-user systems and appended special database which is composed of thermochemical formation data and reaction data. In this report, the detailed specification of our system design is described and the operating methods are illustrated as a user's manual. Also, expanding current system, or interfacing between this system and other system, this report is very useful as a reference. (Author). 10 refs., 18 tabs., 46 fig

Conceptual structure design of experimental facility for advanced spent fuel conditioning process

International Nuclear Information System (INIS)

Joo, J. S.; Koo, J. H.; Jung, W. M.; Jo, I. J.; Kook, D. H.; Yoo, K. S.

2003-01-01

A study on the advanced spent fuel conditioning process (ACP) is carring out for the effective management of spent fuels of domestic nuclear power plants. This study presents basic shielding design, modification of IMEF's reserve hot cell facility which reserved for future usage, conceptual and structural architecture design of ACP hot cell and its contents, etc. considering the characteristics of ACP. The results of this study will be used for the basic and detail design of ACP demonstration facility, and utilized as basic data for the safety evaluation as essential data for the licensing of the ACP facility

Technology readiness levels for advanced nuclear fuels and materials development

Energy Technology Data Exchange (ETDEWEB)

Carmack, W.J., E-mail: jon.carmack@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States); Braase, L.A.; Wigeland, R.A. [Idaho National Laboratory, Idaho Falls, ID (United States); Todosow, M. [Brookhaven National Laboratory, Upton, NY (United States)

2017-03-15

Highlights: • Definition of nuclear fuels system technology readiness level. • Identification of evaluation criteria for nuclear fuel system TRLs. • Application of TRLs to fuel systems. - Abstract: The Technology Readiness process quantitatively assesses the maturity of a given technology. The National Aeronautics and Space Administration (NASA) pioneered the process in the 1980s to inform the development and deployment of new systems for space applications. The process was subsequently adopted by the Department of Defense (DoD) to develop and deploy new technology and systems for defense applications. It was also adopted by the Department of Energy (DOE) to evaluate the maturity of new technologies in major construction projects. Advanced nuclear fuels and materials development is needed to improve the performance and safety of current and advanced reactors, and ultimately close the nuclear fuel cycle. Because deployment of new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the assessment process to advanced fuel development is useful as a management, communication, and tracking tool. This article provides definition of technology readiness levels (TRLs) for nuclear fuel technology as well as selected examples regarding the methods by which TRLs are currently used to assess the maturity of nuclear fuels and materials under development in the DOE Fuel Cycle Research and Development (FCRD) Program within the Advanced Fuels Campaign (AFC).

Advanced methods of solid oxide fuel cell modeling

CERN Document Server

Milewski, Jaroslaw; Santarelli, Massimo; Leone, Pierluigi

2011-01-01

Fuel cells are widely regarded as the future of the power and transportation industries. Intensive research in this area now requires new methods of fuel cell operation modeling and cell design. Typical mathematical models are based on the physical process description of fuel cells and require a detailed knowledge of the microscopic properties that govern both chemical and electrochemical reactions. ""Advanced Methods of Solid Oxide Fuel Cell Modeling"" proposes the alternative methodology of generalized artificial neural networks (ANN) solid oxide fuel cell (SOFC) modeling. ""Advanced Methods

Evolutionary developments of advanced PWR nuclear fuels and cladding materials

International Nuclear Information System (INIS)

Kim, Kyu-Tae

2013-01-01

Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10 −6 on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure

Development of advanced spent fuel management process

International Nuclear Information System (INIS)

Shin, Young Joon; Cho, S. H.; You, G. S.

2001-04-01

Currently, the economic advantage of any known approach to the back end fuel cycle of a nuclear power reactor has not been well established. Thus the long term storage of the spent fuel in a safe manner is one of the important issues to be resolved in countries where the nuclear power has a relatively heavy weight in power production of that country. At KAERI, as a solution to this particular issue midterm storage of the spent fuel, an alternative approach has been developed. This approach includes the decladding and pulverization process of the spent PWR fuel rod, the reducing process from the uranium oxide to a metallic uranium powder using Li metal in a LiCl salt, the continuous casting process of the reduced metal, and the recovery process of Li from mixed salts by the electrolysis. We conducted the laboratory scale tests of each processes for the technical feasibility and determination for the operational conditions for this approach. Also, we performed the theoretical safety analysis and conducted integral tests for the equipment integration through the Mock-up facility with non-radioactive samples. There were no major issues in the approach, however, material incompatibility of the alkaline metal and oxide in a salt at a high temperature and the reactor that contains the salt became a show stopper of the process. Also the difficulty of the clear separation of the salt with metals reduced from the oxide became a major issue

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.

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.

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.

Radiation-resistant requirements analysis of device and control component for advanced spent fuel management process

Energy Technology Data Exchange (ETDEWEB)

Song, Tai Gil; Park, G. Y.; Kim, S. Y.; Lee, J. Y.; Kim, S. H.; Yoon, J. S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-02-01

It is known that high levels of radiation can cause significant damage by altering the properties of materials. A practical understanding of the effects of radiation - how radiation affects various types of materials and components - is required to design equipment to operate reliably in a gamma radiation environment. When designing equipment to operate in a high gamma radiation environment, such as will be present in a nuclear spent fuel handling facility, several important steps should be followed. In order to active test of the advanced spent fuel management process, the radiation-resistant analysis of the device and control component for active test which is concerned about the radiation environment is conducted. Also the system design process is analysis and reviewed. In the foreign literature, 'threshold' values are generally reported. the threshold values are normally the dose required to begin degradation in a particular material property. The radiation effect analysis for the device of vol-oxidation and metalization, which are main device for the advanced spent fuel management process, is performed by the SCALE 4.4 code. 5 refs., 4 figs., 13 tabs. (Author)

Introducing advanced nuclear fuel cycles in Canada

International Nuclear Information System (INIS)

Duret, M.F.

1978-05-01

The ability of several different advanced fuel cycles to provide energy for a range of energy growth scenarios has been examined for a few special situations of interest in Canada. Plutonium generated from the CANDU-PHW operating on natural uranium is used to initiate advanced fuel cycles in the year 2000. The four fuel cycles compared are: 1) natural uranium in the CANDU-PHW; 2) high burnup thorium cycle in the CANDU-PHW; 3) self-sufficient thorium cycle in the CANDU-PHW; 4) plutonium-uranium cycle in a fast breeder reactor. The general features of the results are quite clear. While any plutonium generated prior to the introduction of the advanced fuel cycle remains, system requirements for natural uranium for each of the advanced fuel cycles are the same and are governed by the rate at which plants operating on natural uranium can be retired. When the accumulated plutonium inventory has been entirely used, natural uranium is again required to provide inventory for the advanced fuel cycle reactors. The time interval during which no uranium is required varies only from about 25 to 40 years for both thorium cycles, depending primarily on the energy growth rate. The breeder does not require the entire plutonium inventory produced and so would call for less processing of fuel from the PHW reactors. (author)

Development of the advanced CANDU technology -Development of CANDU advanced fuel fabrication technology-

Energy Technology Data Exchange (ETDEWEB)

Choi, Chang Bum; Park, Choon Hoh; Park, Chul Joo; Kwon, Woo Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

This project is carrying out jointly with AECL to develop CANFLEX fuel which can enhance reactor safety, fuel economy and can be used with various fuel cycles (natural U, slightly enriched U, other advanced fuel). The final goal of this research is to load the CANFLEX fuel in commercial CANDU reactor for demonstration irradiation. The annual portion of research activities performed during this year are followings ; The detail design of CANFLEX-NU fuel was determined. Based on this design, various fabrication drawings and process specifications were revised. The seventeen CANFLEX-NU fuel bundles for reactivity test in ZED-2 and out-pile test, two CANFLEX-SEU fuel bundles for demo-irradiation in NRU were fabricated. Advanced tack welding machine was designed and sequence control software of automatic assembly welder was developed. The basic researches related to fabrication processes, such as weld evaluation by ECT, effect of additives in UO{sub 2}, thermal stabilities of Zr based metallic glasses, were curried out. 51 figs, 22 tabs, 42 refs. (Author).

Experimental research subject and renovation of chemical processing facility (CPF) for advanced fast reactor fuel reprocessing technology development

International Nuclear Information System (INIS)

Koyama, Tomozo; Shinozaki, Tadahiro; Nomura, Kazunori; Koma, Yoshikazu; Miyachi, Shigehiko; Ichige, Yoshiaki; Kobayashi, Tsuguyuki; Nemoto, Shin-ichi

2002-01-01

In order to enhance economical efficiency, environmental impact and nuclear nonproliferation resistance, the Advanced Reprocessing Technology, such as simplification and optimization of process, and applicability evaluation of the innovative technology that was not adopted up to now, has been developed for the reprocessing of the irradiated fuel taken out from a fast reactor. Renovation of the hot cell interior equipments, establishment and updating of glove boxes, installation of various analytical equipments, etc. in the Chemical Processing Facility (CPF) was done to utilize the CPF more positivity which is the center of the experimental field, where actual fuel can be used, for research and development towards establishment of the Advanced Reprocessing Technology development. The hot trials using the irradiated fuel pins of the experimental fast reactor 'JOYO' for studies on improved aqueous reprocessing technology, MA separation technology, dry process technology, etc. are scheduled to be carried out with these new equipments. (author)

Performance Evaluation of the Neutron Coincidence Counter for the Advanced Spent Fuel Conditioning Process

International Nuclear Information System (INIS)

Lee, S.Y.; Li, T.K.; Menlove, Howard O.; Kim, H.D.; Ko, W.I.; Park, S.W.

2005-01-01

The Advanced Spent Fuel Conditioning Process (ACP) is a pyrochemical dry reprocessing technique to convert oxide-type spent nuclear fuel into a metallic form. The Korea Atomic Energy Research Institute (KAERI) has been developing this technology for the purpose of spent fuel management and is planning to perform a lab-scale demonstration in 2006. With this technology, a significant reduction of the volume and heat load of spent fuel is expected, which could decrease the burden of safety and economics. In this study, MCNPX code calculations were carried out to estimate the performance of a neutron coincidence counter designed for measruement of the process materials in the pilot-scale ACP facility. To verify the design requirement, the singles and doubles counting rates of the detectors were simulated with the latest coincidence capability of the MCNPX code. Then, the precision of the coincidence measurements were evaluated on various process materials from the ACP. It was verified that the performance of the neutron coincidence counter could meet the design criteria for all samples in the ACP, and the material accounting system for the pilot-scale ACP facility could meet the IAEA safeguards goals.

Advanced nuclear fuel cycles activities in IAEA

International Nuclear Information System (INIS)

Nawada, H.P.; Ganguly, C.

2007-01-01

MA losses in pyrochemical processes, MA fuel and target, coated particle fuel, MA property data bank, and IMF fuel as well as highlights of the recent technical meeting on liquid-metal-cooled reactor fuel cycle. In addition, the necessity for determination of accurate high temperature thermodynamic data of minor actinides for developing the advanced nuclear fuel cycle would be discussed. (authors)

Advanced Diagnostics in Oxy-Fuel Combustion Processes

DEFF Research Database (Denmark)

Brix, Jacob; Toftegaard, Maja Bøg; Clausen, Sønnik

This report sums up the findings in PSO-project 010069, “Advanced Diagnostics in Oxy- Fuel Combustion Processes”. Three areas of optic diagnostics are covered in this work: - FTIR measurements in a 30 kW swirl burner. - IR measurements in a 30 kW swirl burner. - IR measurements in a laboratory...... technique was an invaluable tool in the discussion of data obtained by gas analysis, and it allowed for estimation of combustion times in O2/CO2 where the high CO2 concentration prevents the use of the carbon mass balance for that purpose. During the project the data have been presented at a conference......, formed the basis of a publication and it is part of two PhD dissertations. The name of the conference the journal and the dissertations are listed below. - Joint Meeting of the Scandinavian-Nordic and French Sections of the Combustion Institute, Combustion of Char Particles under Oxy-Fuel Conditions...

Advanced fuel technology and performance

International Nuclear Information System (INIS)

1985-10-01

The purpose of the Advisory Group Meeting on Advanced Fuel Technology and Performance was to review the experience of advanced fuel fabrication technology, its performance, peculiarities of the back-end of the nuclear fuel cycle with regard to all types of reactors and to outline the future trends. As a result of the meeting recommendations were made for the future conduct of work on advanced fuel technology and performance. A separate abstract was prepared for each of the 20 papers in this issue

Recent advances in the development of a cobalt dicarbollide based solvent extraction process for the separation of Cs and Sr from spent fuel

International Nuclear Information System (INIS)

Law, Jack D.; Todd, Terry A.; Peterman, D.R.; Herbst, R.S.; Tillotson, R.D.

2004-01-01

As part of the Advanced Fuel Cycle Initiative (AFCI), a chlorinated cobalt dicarbollide (CCD)/polyethylene glycol (PEG) based solvent extraction process is being developed for the separation of Cs and Sr from leached spent light water reactor (LWR) fuel. The separation of Cs and Sr would significantly reduce the short-term heat generation of spent nuclear fuel requiring geological disposal. Recent advances in the development of a CCD/PEG process will be presented. The data presented will include acid dependency data, results of batch contact testing using simulant feeds traced with 137 Cs, 90 Sr and 241 Am as well as results of testing to evaluate extractant composition. The impacts of other separation process in an advanced aqueous separation flow sheet on the effectiveness of the CCD/PEG process will be detailed. (authors)

Advanced LWR Nuclear Fuel Cladding Development

International Nuclear Information System (INIS)

Bragg-Sitton, S.; Griffith, G.

2012-01-01

The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R and D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental enhancements are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction to allow improved fuel economy via power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an 'accident tolerant' fuel system that would offer improved coping time under accident scenarios. In a staged development approach, the LWRS program will engage stakeholders throughout the development process to ensure commercial viability of the investigated technologies. Applying minimum performance criteria, several of the top-ranked materials and fabrication concepts will undergo a rigorous series of mechanical, thermal and chemical characterization tests to better define their properties and operating potential in a relatively low-cost, nonnuclear test series. A reduced number of options will be recommended for test rodlet fabrication and in-pile nuclear testing under steady-state, transient and accident conditions. (author)

Advanced Fuels Campaign FY 2010 Accomplishments Report

Energy Technology Data Exchange (ETDEWEB)

Lori Braase

2010-12-01

The Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) Accomplishment Report documents the high-level research and development results achieved in fiscal year 2010. The AFC program has been given responsibility to develop advanced fuel technologies for the Department of Energy (DOE) using a science-based approach focusing on developing a microstructural understanding of nuclear fuels and materials. The science-based approach combines theory, experiments, and multi-scale modeling and simulation aimed at a fundamental understanding of the fuel fabrication processes and fuel and clad performance under irradiation. The scope of the AFC includes evaluation and development of multiple fuel forms to support the three fuel cycle options described in the Sustainable Fuel Cycle Implementation Plan4: Once-Through Cycle, Modified-Open Cycle, and Continuous Recycle. The word “fuel” is used generically to include fuels, targets, and their associated cladding materials. This document includes a brief overview of the management and integration activities; but is primarily focused on the technical accomplishments for FY-10. Each technical section provides a high level overview of the activity, results, technical points of contact, and applicable references.

Advanced Fuels Campaign FY 2014 Accomplishments Report

Energy Technology Data Exchange (ETDEWEB)

Braase, Lori [Idaho National Lab. (INL), Idaho Falls, ID (United States). INL Systems Analyses; May, W. Edgar [Idaho National Lab. (INL), Idaho Falls, ID (United States). INL Systems Analyses

2014-10-01

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.” In support of the Fuel Cycle Research and Development (FCRD) program, AFC is responsible for developing advanced fuels technologies to support the various fuel cycle options defined in the Department of Energy (DOE) Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. 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. The modeling and simulation activities for fuel performance are carried out under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, which is closely coordinated with AFC. In this report, the word “fuel” is used generically to include fuels, targets, and their associated cladding materials. R&D of light water reactor (LWR) fuels with enhanced accident tolerance is also conducted by AFC. These fuel systems are designed to achieve significantly higher fuel and plant performance to allow operation to significantly higher burnup, and to provide enhanced safety during design basis and beyond design basis accident conditions. The overarching goal is to develop advanced nuclear fuels and materials that are robust, have high performance capability, and are more tolerant to

Development of challengeable reprocessing and fuel fabrication technologies for advanced fast reactor fuel cycle

International Nuclear Information System (INIS)

Nomura, S.; Aoshima, T.; Myochin, M.

2001-01-01

R and D in the next five years in Feasibility Study Phase-2 are focused on selected key technologies for the advanced fuel cycle. These are the reference technology of simplified aqueous extraction and fuel pellet short process based on the oxide fuel and the innovative technology of oxide-electrowinning and metal- electrorefining process and their direct particle/metal fuel fabrication methods in a hot cell. Automatic and remote handling system operation in both reprocessing and fuel manufacturing can handle MA and LLFP concurrently with Pu and U attaining the highest recovery and an accurate accountability of these materials. (author)

Overview of the CANDU fuel handling system for advanced fuel cycles

International Nuclear Information System (INIS)

Koivisto, D.J.; Brown, D.R.

1997-01-01

Because of its neutron economies and on-power re-fuelling capabilities the CANDU system is ideally suited for implementing advanced fuel cycles because it can be adapted to burn these alternative fuels without major changes to the reactor. The fuel handling system is adaptable to implement advanced fuel cycles with some minor changes. Each individual advanced fuel cycle imposes some new set of special requirements on the fuel handling system that is different from the requirements usually encountered in handling the traditional natural uranium fuel. These changes are minor from an overall plant point of view but will require some interesting design and operating changes to the fuel handling system. Some preliminary conceptual design has been done on the fuel handling system in support of these fuel cycles. Some fuel handling details were studies in depth for some of the advanced fuel cycles. This paper provides an overview of the concepts and design challenges. (author)

Preliminary assessment of safeguardability on the concepture design of advanced spent fuel conditioning process

Energy Technology Data Exchange (ETDEWEB)

Lee, Sang Yoon; Ha, Jang Ho; Ko, Won Il; Song, Dae Yong; Kim, Ho Dong

2003-04-01

In this report, a preliminary study on the safeguardability of ACP (Advanced spent fuel Conditioning Process) was conducted with Los Alamos National Laboratory. The proposed ACP concept is an electrometallurgical treatment technique to convert oxide-type spent nuclear fuels into metal forms, which can achieve significant reduction of the volume and heat load of spent fuel to be stored and disposed of. For the safeguardability analysis of the ACP facility, sub-processes and their KMPs (Key Measurement Points) were defined first, and then their material flows were analyzed. Finally, the standard deviation of the Inventory Difference (ID) value of the facility was estimated with assumption by assuming international target values for the uncertainty of measurement methods and their uncertainty. From the preliminary calculation, we concluded that if the assumptions regarding measurement instruments can be achieved in a safeguards system for the ACP facility, the safeguards goals of International Atomic Energy Agency (IAEA) could be met. In the second phase of this study, further study on sensitivity analyses considering various factors such as measurement errors, facility capacities, MBA periods etc. may be needed.

Preliminary assessment of safeguardability on the concepture design of advanced spent fuel conditioning process

International Nuclear Information System (INIS)

Lee, Sang Yoon; Ha, Jang Ho; Ko, Won Il; Song, Dae Yong; Kim, Ho Dong

2003-04-01

In this report, a preliminary study on the safeguardability of ACP (Advanced spent fuel Conditioning Process) was conducted with Los Alamos National Laboratory. The proposed ACP concept is an electrometallurgical treatment technique to convert oxide-type spent nuclear fuels into metal forms, which can achieve significant reduction of the volume and heat load of spent fuel to be stored and disposed of. For the safeguardability analysis of the ACP facility, sub-processes and their KMPs (Key Measurement Points) were defined first, and then their material flows were analyzed. Finally, the standard deviation of the Inventory Difference (ID) value of the facility was estimated with assumption by assuming international target values for the uncertainty of measurement methods and their uncertainty. From the preliminary calculation, we concluded that if the assumptions regarding measurement instruments can be achieved in a safeguards system for the ACP facility, the safeguards goals of International Atomic Energy Agency (IAEA) could be met. In the second phase of this study, further study on sensitivity analyses considering various factors such as measurement errors, facility capacities, MBA periods etc. may be needed

Advanced fuel cycles in CANDU reactors

International Nuclear Information System (INIS)

Green, R.E.; Boczar, P.G.

1990-04-01

This paper re-examines the rationale for advanced nuclear fuel cycles in general, and for CANDU advanced fuel cycles in particular. The traditional resource-related arguments for more uranium nuclear fuel cycles are currently clouded by record-low prices for uranium. However, the total known conventional uranium resources can support projected uranium requirements for only another 50 years or so, less if a major revival of the nuclear option occurs as part of the solution to the world's environmental problems. While the extent of the uranium resource in the earth's crust and oceans is very large, uncertainty in the availability and price of uranium is the prime resource-related motivation for advanced fuel cycles. There are other important reasons for pursuing advanced fuel cycles. The three R's of the environmental movement, reduce, recycle, reuse, can be achieved in nuclear energy production through the employment of advanced fuel cycles. The adoption of more uranium-conserving fuel cycles would reduce the amount of uranium which needs to be mined, and the environmental impact of that mining. Environmental concerns over the back end of the fuel cycle can be mitigated as well. Higher fuel burnup reduces the volume of spent fuels which needs to be disposed of. The transmutation of actinides and long-lived fission products into short-lived fission products would reduce the radiological hazard of the waste from thousands to hundreds of years. Recycling of uranium and/or plutonium in spent fuel reuses valuable fissile material, leaving only true waste to be disposed of. Advanced fuel cycles have an economical benefit as well, enabling a ceiling to be put on fuel cycle costs, which are

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

International Nuclear Information System (INIS)

Johnson, E.R.; Best, R.E.

2009-01-01

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 of the

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

Advanced Technology and Alternative Fuel Vehicles

International Nuclear Information System (INIS)

Tuttle, J.

2001-01-01

This fact sheet provides a basic overview of today's alternative fuel choices--including biofuels, biodiesel, electricity, and hydrogen--alternative fuel vehicles, and advanced vehicle technology, such as hybrid electric vehicles, fuel cells and advanced drive trains

Task 3.3: Warm Syngas Cleanup and Catalytic Processes for Syngas Conversion to Fuels Subtask 3: Advanced Syngas Conversion to Fuels

Energy Technology Data Exchange (ETDEWEB)

Lebarbier Dagel, Vanessa M.; Li, J.; Taylor, Charles E.; Wang, Yong; Dagle, Robert A.; Deshmane, Chinmay A.; Bao, Xinhe

2014-03-31

activity was to develop methods and enabling materials for syngas conversion to SNG with readily CO2 separation. Suitable methanation catalyst and CO2 sorbent materials were developed. Successful proof-of-concept for the combined reaction-sorption process was demonstrated, which culminated in a research publication. With successful demonstration, a decision was made to switch focus to an area of fuels research of more interest to all three research institutions (CAS-NETL-PNNL). Syngas-to-Hydrocarbon Fuels through Higher Alcohol Intermediates There are two types of processes in syngas conversion to fuels that are attracting R&D interest: 1) syngas conversion to mixed alcohols; and 2) syngas conversion to gasoline via the methanol-to-gasoline process developed by Exxon-Mobil in the 1970s. The focus of this task was to develop a one-step conversion technology by effectively incorporating both processes, which is expected to reduce the capital and operational cost associated with the conversion of coal-derived syngas to liquid fuels. It should be noted that this work did not further study the classic Fischer-Tropsch reaction pathway. Rather, we focused on the studies for unique catalyst pathways that involve the direct liquid fuel synthesis enabled by oxygenated intermediates. Recent advances made in the area of higher alcohol synthesis including the novel catalytic composite materials recently developed by CAS using base metal catalysts were used.

Advanced Fuels Campaign FY 2015 Accomplishments Report

Energy Technology Data Exchange (ETDEWEB)

Braase, Lori Ann [Idaho National Lab. (INL), Idaho Falls, ID (United States); Carmack, William Jonathan [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-10-29

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 report is a compilation of technical accomplishment summaries for FY-15. Emphasis is on advanced accident-tolerant LWR fuel systems, advanced transmutation fuels technologies, and capability development.

Manufacturing technology and process for BWR fuel

International Nuclear Information System (INIS)

Kato, Shigeru

1996-01-01

Following recent advanced technologies, processes and requests of the design changes of BWR fuel, Nuclear Fuel Industries, Ltd. (NFI) has upgraded the manufacturing technology and honed its own skills to complete its brand-new automated facility in Tokai in the latter half of 1980's. The plant uses various forms of automation throughout the manufacturing process: the acceptance of uranium dioxide powder, pelletizing, fuel rod assembling, fuel bundle assembling and shipment. All processes are well computerized and linked together to establish the integrated control system with three levels of Production and Quality Control, Process Control and Process Automation. This multi-level system plays an important role in the quality assurance system which generates the highest quality of fuels and other benefits. (author)

LOFT advanced fuel rod instrumentation development

International Nuclear Information System (INIS)

Billeter, T.R.; Brown, R.L.; Chan, A.I.Y.; Day, C.K.; Meyers, S.C.; Sheen, E.M.; Stringer, J.L.

1978-01-01

Advanced fuel rod instrumentation for the Loss of Fluid Test (LOFT) reactor is being developed by the Hanford Engineering Development Laboratory for the Nuclear Regulatory Commission. This effort calls for development of sensors to measure fuel rod axial motion, fuel centerline temperature (to 2200 0 C), fuel rod plenum gas pressure (to 2500 psig), and plenum gas temperature (to 1500 0 F). A parallel test and evaluation of several modified commercial sensors was undertaken and will result in commercial availability of the final qualified sensors. Necessary test facilities were prepared for the development and evaluation effort. Tests to date indicate a three coil Linear Variable Differential Transformer (LVDT), operated from temperature compensating signal source and processing electronics, will meet the desired requirements

State-of-the-art Report on Innovative Fuels for Advanced Nuclear Systems

International Nuclear Information System (INIS)

Chauvin, N.; Minato, K.; Ogata, T.; Lee, C.B.; Pouchon, M.A.; Pasamehmetoglu, K.O.; Choi, Y.J.; Kennedy, J.R.; Massara, S.; Cornet, S.; ); Sommers, J.; ); McClellan, K.

2014-01-01

Development of innovative fuels such as homogeneous and heterogeneous fuels, ADS fuels, and oxide, metal, nitride and carbide fuels is an important stage in the implementation process of advanced nuclear systems. Several national and international R and D programmes are investigating minor actinide-bearing fuels due to their ability to help reduce the radiotoxicity of spent fuel and therefore decrease the burden on geological repositories. Minor actinides can be converted into a suitable fuel form for irradiation in reactor systems where they are transmuted into fission products with a significantly shorter half-life. This report compares recent studies of fuels containing minor actinides for use in advanced nuclear systems. The studies review different fuels for several types of advanced reactors by examining various technical issues associated with fabrication, characterisation, irradiation performance, design and safety criteria, as well as technical maturity. (authors)

Development of CANDU advanced fuel bundle

International Nuclear Information System (INIS)

Suk, H. C.; Hwang, W.; Rhee, B. W.; Jung, S. H.; Chung, C. H.

1992-05-01

This research project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle (so-called, CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactor for 1996 and 1997, and consequently will be used in the existing and future CANDU reactors in Korea. The research activities during this year include the detail design of CANFLEX fuel with natural enriched uranium (CANFLEX-NU). Based on this design, CANFLEX fuel was mocked up. Out-of-pile hydraulic scoping tests were conducted with the fuel in the CANDU Cold Test Loop to investigate the condition under which maximum pressure drop occurs and the maximum value of the bundle pressure drop. (Author)

Effect of advanced fuel cycles on waste management policies

International Nuclear Information System (INIS)

Cavedon, J.M.; Haapalehto, T.

2005-01-01

The study aims at analysing a range of future fuel cycle options from the perspective of their impact on waste repository demand and specification. The study would focus on: Assessment of the characteristics of radioactive wastes arising from advanced nuclear fuel cycle options, repository performance analysis studies using source terms for waste arising from such advanced nuclear fuel cycles, identification of new options for waste management and disposal. Three families of fuel cycles having increasing recycling capabilities are assessed. Each cycle is composed of waste generating and management processes. Examples of waste generating processes are fuel factories (7 types) and reprocessing plants (7 types). Packaging and conditioning plants (7) and disposal facilities are examples of waste management processes. The characteristic of all these processes have been described and then total waste flows are summarised. In order to simplify the situation, three waste categories have been defined based on the IAEA definitions in order to emphasize the major effects of different types of waste. These categories are: short-life waste for surface or sub-surface disposal, long-life low heat producing waste for geological disposal, high-level waste for geological disposal. The feasibilities of the fuel cycles are compared in terms of economics, primary resource consumption and amount of waste generated. The effect of high-level waste composition for the repository performance is one of the tools in these comparisons. The results of this will be published as an NEA publication before the end of 2005. (authors)

Feasibility Study for Monitoring Actinide Elements in Process Materials Using FO-LIBS at Advanced spent fuel Conditioning Process Facility

Energy Technology Data Exchange (ETDEWEB)

Han, Bo-Young; Choi, Daewoong; Park, Se Hwan; Kim, Ho-Dong [Nonproliferation System Research Division, Korea Atomic Energy Research Institute, Daejeon, 305-353 (Korea, Republic of); Dae, Dongsun [Department of Chemistry, Mokpo National University, Jeonnam 534-729 (Korea, Republic of); Whitehouse, Andrew I. [Applied Photonics Ltd., Unit 8 Carleton Business Park, Skipton, North Yorkshire BD23 2DE (United Kingdom)

2015-07-01

Korea Atomic Energy Research Institute (KAERI) have been developing the design and deployment methodology of Laser- Induced Breakdown Spectroscopy (LIBS) instrument for safeguards application within the argon hot cell environment at Advanced spent fuel Conditioning Process Facility (ACPF), where ACPF is a facility being refurbished for the laboratory-scaled demonstration of advanced spent fuel conditioning process. LIBS is an analysis technology used to measure the emission spectra of excited elements in the local plasma of a target material induced by a laser. The spectra measured by LIBS are analyzed to verify the quality and quantity of the specific element in the target matrix. Recently LIBS has been recognized as a promising technology for safeguards purposes in terms of several advantages including a simple sample preparation and in-situ analysis capability. In particular, a feasibility study of LIBS to remotely monitor the nuclear material in a high radiation environment has been carried out for supporting the IAEA safeguards implementation. Fiber-Optic LIBS (FO-LIBS) deployment was proposed by Applied Photonics Ltd because the use of fiber optics had benefited applications of LIBS by delivering the laser energy to the target and by collecting the plasma light. The design of FO-LIBS instrument for the measurement of actinides in the spent fuel and high temperature molten salt at ACPF had been developed in cooperation with Applied Photonics Ltd. FO-LIBS has some advantages as followings: the detectable plasma light wavelength range is not limited by the optical properties of the thick lead-glass shield window and the potential risk of laser damage to the lead-glass shield window is not considered. The remote LIBS instrument had been installed at ACPF and then the feasibility study for monitoring actinide elements such as uranium, plutonium, and curium in process materials has been carried out. (authors)

Feasibility Study for Monitoring Actinide Elements in Process Materials Using FO-LIBS at Advanced spent fuel Conditioning Process Facility

International Nuclear Information System (INIS)

Han, Bo-Young; Choi, Daewoong; Park, Se Hwan; Kim, Ho-Dong; Dae, Dongsun; Whitehouse, Andrew I.

2015-01-01

Korea Atomic Energy Research Institute (KAERI) have been developing the design and deployment methodology of Laser- Induced Breakdown Spectroscopy (LIBS) instrument for safeguards application within the argon hot cell environment at Advanced spent fuel Conditioning Process Facility (ACPF), where ACPF is a facility being refurbished for the laboratory-scaled demonstration of advanced spent fuel conditioning process. LIBS is an analysis technology used to measure the emission spectra of excited elements in the local plasma of a target material induced by a laser. The spectra measured by LIBS are analyzed to verify the quality and quantity of the specific element in the target matrix. Recently LIBS has been recognized as a promising technology for safeguards purposes in terms of several advantages including a simple sample preparation and in-situ analysis capability. In particular, a feasibility study of LIBS to remotely monitor the nuclear material in a high radiation environment has been carried out for supporting the IAEA safeguards implementation. Fiber-Optic LIBS (FO-LIBS) deployment was proposed by Applied Photonics Ltd because the use of fiber optics had benefited applications of LIBS by delivering the laser energy to the target and by collecting the plasma light. The design of FO-LIBS instrument for the measurement of actinides in the spent fuel and high temperature molten salt at ACPF had been developed in cooperation with Applied Photonics Ltd. FO-LIBS has some advantages as followings: the detectable plasma light wavelength range is not limited by the optical properties of the thick lead-glass shield window and the potential risk of laser damage to the lead-glass shield window is not considered. The remote LIBS instrument had been installed at ACPF and then the feasibility study for monitoring actinide elements such as uranium, plutonium, and curium in process materials has been carried out. (authors)

Advances in nuclear fuel technology. 3. Development of advanced nuclear fuel recycle systems

International Nuclear Information System (INIS)

Arie, Kazuo; Abe, Tomoyuki; Arai, Yasuo

2002-01-01

Fast breeder reactor (FBR) cycle technology has a technical characteristics flexibly easy to apply to diverse fuel compositions such as plutonium, minor actinides, and so on and fuel configurations. By using this characteristics, various feasibilities on effective application of uranium resources based on breeding of uranium of plutonium for original mission of FBR, contribution to radioactive wastes problems based on amounts reduction of transuranium elements (TRU) in high level radioactive wastes, upgrading of nuclear diffusion resistance, extremely upgrading of economical efficiency, and so on. In this paper, were introduced from these viewpoints, on practice strategy survey study on FBR cycle performed by cooperation of the Japan Nuclear Cycle Development Institute (JNC) with electric business companies and so on, and on technical development on advanced nuclear fuel recycle systems carried out at the Central Research Institute of Electric Power Industry, Japan Atomic Energy Research Institute, and so on. Here were explained under a vision on new type of fuels such as nitride fuels, metal fuels, and so on as well as oxide fuels, a new recycle system making possible to use actinides except uranium and plutonium, an 'advanced nuclear fuel cycle technology', containing improvement of conventional wet Purex method reprocessing technology, fuel manufacturing technology, and so on. (G.K.)

Selection and development of advanced nuclear fuel products

International Nuclear Information System (INIS)

Stucker, David L.; Miller, Richard S.; Arnsberger, Peter L.

2004-01-01

The highly competitive international marketplace requires a continuing product development commitment, short development cycle times and timely, on-target product development to assure customer satisfaction and continuing business. Westinghouse has maintained its leadership position within the nuclear fuel industry with continuous developments and improvements to fuel assembly materials and design. This paper presents a discussion of the processes used by Westinghouse in the selection and refinement of advanced concepts for deployment in the highly competitive US and international nuclear fuel fabrication marketplace. (author)

Powder handling for automated fuel processing

International Nuclear Information System (INIS)

Frederickson, J.R.; Eschenbaum, R.C.; Goldmann, L.H.

1989-01-01

Installation of the Secure Automated Fabrication (SAF) line has been completed. It is located in the Fuel Cycle Plant (FCP) at the Department of Energy's (DOE) Hanford site near Richland, Washington. The SAF line was designed to fabricate advanced reactor fuel pellets and assemble fuel pins by automated, remote operation. This paper describes powder handling equipment and techniques utilized for automated powder processing and powder conditioning systems in this line. 9 figs

Safety evaluation report of hot cell facilities for demonstration of advanced spent fuel conditioning process

International Nuclear Information System (INIS)

You, Gil Sung; Choung, W. M.; Ku, J. H.; Cho, I. J.; Kook, D. H.; Park, S. W.; Bek, S. Y.; Lee, E. P.

2004-10-01

The advanced spent fuel conditioning process(ACP) proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel. In the next phase(2004∼2006), the hot test will be carried out for verification of the ACP in a laboratory scale. For the hot test, the hot cell facilities of α- type and auxiliary facilities are required essentially for safe handling of high radioactive materials. As the hot cell facilities for demonstration of the ACP, a existing hot cell of β- type will be refurbished to minimize construction expenditures of hot cell facility. Up to now, the detail design of hot cell facilities and process were completed, and the safety analysis was performed to substantiate secure of conservative safety. The design data were submitted for licensing which was necessary for construction and operation of hot cell facilities. The safety investigation of KINS on hot cell facilities was completed, and the license for construction and operation of hot cell facilities was acquired already from MOST. In this report, the safety analysis report submitted to KINS was summarized. And also, the questionnaires issued from KINS and answers of KAERI in process of safety investigation were described in detail

Fuel quality processing study, volume 1

Science.gov (United States)

Ohara, J. B.; Bela, A.; Jentz, N. E.; Syverson, H. T.; Klumpe, H. W.; Kessler, R. E.; Kotzot, H. T.; Loran, B. L.

1981-01-01

A fuel quality processing study to provide a data base for an intelligent tradeoff between advanced turbine technology and liquid fuel quality, and also, to guide the development of specifications of future synthetic fuels anticipated for use in the time period 1985 to 2000 is given. Four technical performance tests are discussed: on-site pretreating, existing refineries to upgrade fuels, new refineries to upgrade fuels, and data evaluation. The base case refinery is a modern Midwest refinery processing 200,000 BPD of a 60/40 domestic/import petroleum crude mix. The synthetic crudes used for upgrading to marketable products and turbine fuel are shale oil and coal liquids. Of these syncrudes, 50,000 BPD are processed in the existing petroleum refinery, requiring additional process units and reducing petroleum feed, and in a new refinery designed for processing each syncrude to produce gasoline, distillate fuels, resid fuels, and turbine fuel, JPGs and coke. An extensive collection of synfuel properties and upgrading data was prepared for the application of a linear program model to investigate the most economical production slate meeting petroleum product specifications and turbine fuels of various quality grades. Technical and economic projections were developed for 36 scenarios, based on 4 different crude feeds to either modified existing or new refineries operated in 2 different modes to produce 7 differing grades of turbine fuels. A required product selling price of turbine fuel for each processing route was calculated. Procedures and projected economics were developed for on-site treatment of turbine fuel to meet limitations of impurities and emission of pollutants.

Dynamic Simulations of Advanced Fuel Cycles

International Nuclear Information System (INIS)

Piet, Steven J.; Dixon, Brent W.; Jacobson, Jacob J.; Matthern, Gretchen E.; Shropshire, David E.

2011-01-01

Years of performing dynamic simulations of advanced nuclear fuel cycle options provide insights into how they could work and how one might transition from the current once-through fuel cycle. This paper summarizes those insights from the context of the 2005 objectives and goals of the U.S. Advanced Fuel Cycle Initiative (AFCI). Our intent is not to compare options, assess options versus those objectives and goals, nor recommend changes to those objectives and goals. Rather, we organize what we have learned from dynamic simulations in the context of the AFCI objectives for waste management, proliferation resistance, uranium utilization, and economics. Thus, we do not merely describe 'lessons learned' from dynamic simulations but attempt to answer the 'so what' question by using this context. The analyses have been performed using the Verifiable Fuel Cycle Simulation of Nuclear Fuel Cycle Dynamics (VISION). We observe that the 2005 objectives and goals do not address many of the inherently dynamic discriminators among advanced fuel cycle options and transitions thereof.

CANDU-6 fuel optimization for advanced cycles

Energy Technology Data Exchange (ETDEWEB)

St-Aubin, Emmanuel, E-mail: emmanuel.st-aubin@polymtl.ca; Marleau, Guy, E-mail: guy.marleau@polymtl.ca

2015-11-15

Highlights: • New fuel selection process proposed for advanced CANDU cycles. • Full core time-average CANDU modeling with independent refueling and burnup zones. • New time-average fuel optimization method used for discrete on-power refueling. • Performance metrics evaluated for thorium-uranium and thorium-DUPIC cycles. - Abstract: We implement a selection process based on DRAGON and DONJON simulations to identify interesting thorium fuel cycles driven by low-enriched uranium or DUPIC dioxide fuels for CANDU-6 reactors. We also develop a fuel management optimization method based on the physics of discrete on-power refueling and the time-average approach to maximize the economical advantages of the candidates that have been pre-selected using a corrected infinite lattice model. Credible instantaneous states are also defined using a channel age model and simulated to quantify the hot spots amplitude and the departure from criticality with fixed reactivity devices. For the most promising fuels identified using coarse models, optimized 2D cell and 3D reactivity device supercell DRAGON models are then used to generate accurate reactor databases at low computational cost. The application of the selection process to different cycles demonstrates the efficiency of our procedure in identifying the most interesting fuel compositions and refueling options for a CANDU reactor. The results show that using our optimization method one can obtain fuels that achieve a high average exit burnup while respecting the reference cycle safety limits.

Assessment of Startup Fuel Options for the GNEP Advanced Burner Reactor (ABR)

Energy Technology Data Exchange (ETDEWEB)

Jon Carmack (062056); Kemal O. Pasamehmetoglu (103171); David Alberstein

2008-02-01

The Global Nuclear Energy Program (GNEP) includes a program element for the development and construction of an advanced sodium cooled fast reactor to demonstrate the burning (transmutation) of significant quantities of minor actinides obtained from a separations process and fabricated into a transuranic bearing fuel assembly. To demonstrate and qualify transuranic (TRU) fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype is needed. The ABR would necessarily be started up using conventional metal alloy or oxide (U or U, Pu) fuel. Startup fuel is needed for the ABR for the first 2 to 4 core loads of fuel in the ABR. Following start up, a series of advanced TRU bearing fuel assemblies will be irradiated in qualification lead test assemblies in the ABR. There are multiple options for this startup fuel. This report provides a description of the possible startup fuel options as well as possible fabrication alternatives available to the program in the current domestic and international facilities and infrastructure.

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.

Practice and prospect of advanced fuel management and fuel technology application in PWR in China

International Nuclear Information System (INIS)

Xiao Min; Zhang Hong; Ma Cang; Bai Chengfei; Zhou Zhou; Wang Lei; Xiao Xiaojun

2015-01-01

Since Daya Bay nuclear power plant implemented 18-month refueling strategy in 2001, China has completed a series of innovative fuel management and fuel technology projects, including the Ling Ao Advanced Fuel Management (AFM) project (high-burnup quarter core refueling) and the Ningde 18-month refueling project with gadolinium-bearing fuel in initial core. First, this paper gives brief introduction to China's advanced fuel management and fuel technology experience. Second, it introduces practices of the advanced fuel management in China in detail, which mainly focuses on the implementation and progress of the Ningde 18-month refueling project with gadolinium-bearing fuel in initial core. Finally, the paper introduces the practices of advanced fuel technology in China and gives the outlook of the future advanced fuel management and fuel technology in this field. (author)

Property-process relationships in nuclear fuel fabrication

International Nuclear Information System (INIS)

Tikare, V.

2015-01-01

Nuclear fuels are fabricated using many different techniques as they come in a large variety of shapes and compositions. The design and composition of nuclear fuels are predominantly dictated by the engineering requirements necessary for their function in reactors of various designs. Other engineering properties requirements originate from safety and security concerns, and the easy of handling, storing, transporting and disposing of the radioactive materials. In this chapter, the more common of these fuels will be briefly reviewed and the methods used to fabricate them will be presented. The fuels considered in this paper are oxide fuels used in LWRs and FRs, metal fuels in FRs and particulate fuels used in HTGRs. Fabrication of alternative fuel forms and use of standard fuels in alternative reactors will be discussed briefly. The primary motivation to advance fuel fabrication is to improve performance, reduce cost, reduce waste or enhance safety and security of the fuels. To achieve optimal performance, developing models to advance fuel fabrication has to be done in concert with developing fuel performance models. The specific properties and microstructures necessary for improved fuel performance must be identified using fuel performance models, while fuel fabrication models that can determine processing variables to give the desired microstructure and materials properties must be developed. (author)

Advanced fuel fabrication

International Nuclear Information System (INIS)

Bernard, H.

1989-01-01

This paper deals with the fabrication of advanced fuels, such as mixed oxides for Pressurized Water Reactors or mixed nitrides for Fast Breeder Reactors. Although an extensive production experience exists for the mixed oxides used in the FBR, important work is still needed to improve the theoretical and technical knowledge of the production route which will be introduced in the future European facility, named Melox, at Marcoule. Recently, the feasibility of nitride fuel fabrication in existing commercial oxide facilities was demonstrated in France. The process, based on carbothermic reduction of oxides with subsequent comminution of the reaction product, cold pressing and sintering provides (U, Pu)N pellets with characteristics suitable for irradiation testing. Two experiments named NIMPHE 1 and 2 fabricated in collaboration with ITU, Karlsruhe, involve 16 nitride and 2 carbide pins, operating at a linear power of 45 and 73 kW/m with a smear density of 75-80% TD and a high burn-up target of 15 at%. These experiments are currently being irradiated in Phenix, at Marcoule. (orig.)

Advanced Fuels Campaign 2012 Accomplishments

Energy Technology Data Exchange (ETDEWEB)

Not Listed

2012-11-01

The Advanced Fuels Campaign (AFC) under the Fuel Cycle Research and Development (FCRD) program is responsible for developing fuels technologies to support the various fuel cycle options defined in the DOE Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. The fiscal year 2012 (FY 2012) accomplishments are highlighted below. Kemal Pasamehmetoglu is the National Technical Director for AFC.

Advanced fuel development at AECL: What does the future hold for CANDU fuels/fuel cycles?

Energy Technology Data Exchange (ETDEWEB)

Kupferschmidt, W.C.H. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2013-07-01

This paper outlines advanced fuel development at AECL. It discusses expanding the limits of fuel utilization, deploy alternate fuel cycles, increase fuel flexibility, employ recycled fuels; increase safety and reliability, decrease environmental impact and develop proliferation resistant fuel and fuel cycle.

IEA-Advanced Motor Fuels Annual Report 2010

Energy Technology Data Exchange (ETDEWEB)

NONE

2011-12-02

The annual report from the IEA implementing agreement on Advanced Motor Fuels (AMF) describes the agreement, activities, and projects for the year. A section on the global situation for Advanced Motor Fuels includes country reports from each participating AMF member. A status report on each active annex for the agreement is also included, as is a message from the AMF Chairman. Final sections include an Outlook for Advanced Motor Fuels, further information, and a glossary of terms.

PLUS 7TM advanced fuel assembly development program for KSNPs and APR1400

International Nuclear Information System (INIS)

Kim, Kyutae; Stucker, David L.

2002-01-01

KNFC and Westinghouse have recently completed the development of the PLUS 7 TM advanced 16 X 16 fuel assembly for the Korean Standard Nuclear Plants (KSNPs) and the Advanced Power Reactor 1400 (APR 1400). This fuel design utilized the proven advanced design features including mixing vane spacer grids to increase critical heat flux performance, ZIRLO TM advanced materials to enable high-duty, high burnup fuel management and an optimized fuel rod diameter which improves fuel cycle cost while resulting in significant standardization of Korean fuel manufacture. PLUS 7 TM , also includes a patented spacer grid design with conformal fuel rod support designed to provide superior fuel rod wear/fretting resistance while minimizing pressure drop. This paper will present an overview of the PLUS 7 TM fuel assembly development process including a summary of the three-year design and testing program from a mechanical, neutronic, and thermal/hydraulic perspective. The PLUS 7 TM fuel for the KSNPs and the APR1400 reactors results in multi-million dollar per cycle savings in imported enriched uranium product for the Korean nuclear power program with technology specifically developed for Korea by experienced Korean engineers

Equipment system for advanced nuclear fuel development

International Nuclear Information System (INIS)

Kwon, Hyuk Il; Ji, C. G.; Bae, S. O.

2002-11-01

The purpose of the settlement of equipment system for nuclear Fuel Technology Development Facility(FTDF) is to build a seismic designed facility that can accommodate handling of nuclear materials including <20% enriched Uranium and produce HANARO fuel commercially, and also to establish the advanced common research equipment essential for the research on advanced fuel development. For this purpose, this research works were performed for the settlement of radiation protection system and facility special equipment for the FTDF, and the advanced common research equipment for the fuel fabrication and research. As a result, 11 kinds of radiation protection systems such as criticality detection and alarm system, 5 kinds of facility special equipment such as environmental pollution protection system and 5 kinds of common research equipment such as electron-beam welding machine were established. By the settlement of exclusive domestic facility for the research of advanced fuel, the fabrication and supply of HANARO fuel is possible and also can export KAERI-invented centrifugal dispersion fuel materials and its technology to the nations having research reactors in operation. For the future, the utilization of the facility will be expanded to universities, industries and other research institutes

A decade of advances in metallic fuel

International Nuclear Information System (INIS)

Lahm, C.E.; Pahl, R.G.; Porter, D.L.; Tsai, H.; Seidel, B.R.; Batte, G.L.; Dodds, N.E.; Hofman, G.L.; Walters, L.C.

1991-01-01

Significant advances in the understanding of behavior and performance of metallic fuels to high burnup have been achieved over the past four decades. Metallic fuels were the first fuels for liquid-metal-cooled fast reactors (LMR) but in the late 1960's worldwide interest turned toward ceramic fuels before the full potential of metallic fuel could be achieved. Now metallic fuels are recognized as a preferred viable option with regard to safety, integral fuel cycle, waste minimization and deployment economics. This paper reviews the key advances in the last decade and highlights the behavior and performance features which have demonstrated a much greater potential than previously expected

Development of fabrication technology for CANDU advanced fuel -Development of the advanced CANDU technology-

Energy Technology Data Exchange (ETDEWEB)

Choi, Chang Beom; Kim, Hyeong Soo; Kim, Sang Won; Seok, Ho Cheon; Shim, Ki Seop; Byeon, Taek Sang; Jang, Ho Il; Kim, Sang Sik; Choi, Il Kwon; Cho, Dae Sik; Sheo, Seung Won; Lee, Soo Cheol; Kim, Yoon Hoi; Park, Choon Ho; Jeong, Seong Hoon; Kang, Myeong Soo; Park, Kwang Seok; Oh, Hee Kwan; Jang, Hong Seop; Kim, Yang Kon; Shin, Won Cheol; Lee, Do Yeon; Beon, Yeong Cheol; Lee, Sang Uh; Sho, Dal Yeong; Han, Eun Deok; Kim, Bong Soon; Park, Cheol Joo; Lee, Kyu Am; Yeon, Jin Yeong; Choi, Seok Mo; Shon, Jae Moon [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

1994-07-01

The present study is to develop the advanced CANDU fuel fabrication technologies by means of applying the R and D results and experiences gained from localization of mass production technologies of CANDU fuels. The annual portion of this year study includes following: 1. manufacturing of demo-fuel bundles for out-of-pile testing 2. development of technologies for the fabrication and inspection of advanced fuels 3. design and munufacturing of fuel fabrication facilities 4. performance of fundamental studies related to the development of advanced fuel fabrication technology.

Uncertainty Analyses of Advanced Fuel Cycles

International Nuclear Information System (INIS)

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

2008-01-01

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

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.

Safety issues on advanced fuel

International Nuclear Information System (INIS)

Gross, H.; Krebs, W.D.

1998-01-01

In the recent years a general discussion has started whether unsolved safety issues are related to advanced fuel. Advanced fuel is in this context a summary of features like high burnup, improved clad materials, low leakage loading pattern with high peaking factors etc. The design basis accidents RIA and Loca are of special interest for this discussion. From the Siemens point of view RIA is not a safety issue. There are sufficient margins between the enthalpy rise calculated by modern 3D methods and the fuel failures which occurred in RIA simulation tests when the effect of pulse width is taken into account. The evaluation of possible uncertainties for the established Loca criteria (17% equivalent corrosion, 1200 C clad temperature) for high burnup makes sense. But fuel with high burnup has significantly lower peaking factors than fuel with lower burnup. This gives sufficient margin counterbalancing possible uncertainties. In contrast to the above incomplete control rod insertion at higher burnup is potentially a real safety issue. Although Siemens fuel was not affected by the reported incidents they addressed the problem and checked that they have sufficient design margin for their fuel. (orig.) [de

A decade of advances in metallic fuel

International Nuclear Information System (INIS)

Seidel, B.R.; Batte, G.L.; Dodds, N.E.; Hofman, G.L.; Lahm, C.E.; Pahl, R.G.; Porter, D.L.; Tsai, H.; Walters, L.C.

1990-01-01

Significant advances in the understanding of behavior and performance of metallic fuels to high burnup have been achieved over the past four decades. Metallic fuels were the first fuels for liquid-metal-cooled fast reactors (LMR) but in the late 1960s worldwide interest turned toward ceramic fuels before the full potential of metallic fuel could be achieved. Now metallic fuels are recognized as a preferred viable option with regard to safety, integral fuel cycle, waste minimization and deployment economics. This paper reviews the key advances in the last decade and highlights the behavior and performance features which have demonstrated a much greater potential than previously expected. 28 refs., 2 figs., 1 tab

Verification tests for CANDU advanced fuel -Development of the advanced CANDU technology-

International Nuclear Information System (INIS)

Chung, Jang Hwan; Suk, Ho Cheon; Jeong, Moon Ki; Park, Joo Hwan; Jeong, Heung Joon; Jeon, Ji Soo; Kim, Bok Deuk

1994-07-01

This project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle (so-called, CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactor, and consequently will be used in the existing and future CANDU reactors in Korea. The research activities during this year Out-of-pile hydraulic tests for the prototype of CANFLEX bundle was conducted in the CANDU-hot test loop at KAERI. Thermalhydraulic analysis with the assumption of CANFLEX-NU fuel loaded in Wolsong-1 was performed by using thermalhydraulic code, and the thermal margin and T/H compatibility of CANFLEX bundle with existing fuel for CANDU-6 reactor have been evaluated. (Author)

Basic research and industrialization of CANDU advanced fuel

Energy Technology Data Exchange (ETDEWEB)

Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su [and others

2000-04-01

Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and

Basic research and industrialization of CANDU advanced fuel

International Nuclear Information System (INIS)

Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su

2000-04-01

Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and

Future Transient Testing of Advanced Fuels

Energy Technology Data Exchange (ETDEWEB)

Jon Carmack

2009-09-01

The transient in-reactor fuels testing workshop was held on May 4–5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat à l'Énergie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric – Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by

Future Transient Testing of Advanced Fuels

International Nuclear Information System (INIS)

Carmack, Jon

2009-01-01

The transient in-reactor fuels testing workshop was held on May 4-5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat energie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric - Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by the

CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

International Nuclear Information System (INIS)

Chung, Chang Hwan; Chang, S.K.; Hong, S.D.

1999-04-01

A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs

CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

Energy Technology Data Exchange (ETDEWEB)

Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

1999-04-01

A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs.

Design and analysis challenges for advanced nuclear fuel

International Nuclear Information System (INIS)

Klepfer, H.; Abdollahian, D.; Dias, A.; Durston, C.; Eisenhart, L.; Engel, R.; Gilmore, P.; Rank, P.; Kjaer-Pedersen, N.; Sorensen, J.; Yang, R.; Agee, L.

2004-01-01

Significant changes have been incorporated in the light water reactor (LWR) fuel designs now being offered, and advanced fuel designs are currently being developed for the existing and the next generation of reactor designs. These advanced fuel design configurations are intended to offer utilities major economic gains, including: (1) improved fuel characteristics through optimized hydrogen to uranium ratio within the core; (2) increased capacity factor by allowing longer operating cycles, which is implemented by increasing the fuel enrichment and the amount and distribution of burnable poison, gadolinia, boron, or erbium within the fuel assembly to achieve higher discharge burnup; and (3) increased plant power output, if it can be accommodated by the balance of plant, by increasing the power density of the fuel assembly. The authors report here work being done to identify emerging technical issues in support of utility industry evaluations of advanced fuel designs. (author)

Impact of advanced fuel cycles on uncertainty associated with geologic repositories

International Nuclear Information System (INIS)

Rechard, Rob P.; Lee, Joon; Sutton, Mark; Greenberg, Harris R.; Robinson, Bruce A.; Nutt, W. Mark

2013-01-01

This paper provides a qualitative evaluation of the impact of advanced fuel cycles, particularly partition and transmutation of actinides, on the uncertainty associated with geologic disposal. Based on the discussion, advanced fuel cycles, will not materially alter (1) the repository performance (2) the spread in dose results around the mean (3) the modeling effort to include significant features, events, and processes in the performance assessment, or (4) the characterization of uncertainty associated with a geologic disposal system in the regulatory environment of the United States. (authors)

Advanced nuclear fuel cycles and radioactive waste management

International Nuclear Information System (INIS)

2006-01-01

This study analyses a range of advanced nuclear fuel cycle options from the perspective of their effect on radioactive waste management policies. It presents various fuel cycle options which illustrate differences between alternative technologies, but does not purport to cover all foreseeable future fuel cycles. The analysis extends the work carried out in previous studies, assesses the fuel cycles as a whole, including all radioactive waste generated at each step of the cycles, and covers high-level waste repository performance for the different fuel cycles considered. The estimates of quantities and types of waste arising from advanced fuel cycles are based on best available data and experts' judgement. The effects of various advanced fuel cycles on the management of radioactive waste are assessed relative to current technologies and options, using tools such as repository performance analysis and cost studies. (author)

Advanced methods of quality control in nuclear fuel fabrication

International Nuclear Information System (INIS)

Onoufriev, Vladimir

2004-01-01

Under pressure of current economic and electricity market situation utilities implement more demanding fuel utilization schemes including higher burn ups and thermal rates, longer fuel cycles and usage of Mo fuel. Therefore, fuel vendors have recently initiated new R and D programmes aimed at improving fuel quality, design and materials to produce robust and reliable fuel. In the beginning of commercial fuel fabrication, emphasis was given to advancements in Quality Control/Quality Assurance related mainly to product itself. During recent years, emphasis was transferred to improvements in process control and to implementation of overall Total Quality Management (TQM) programmes. In the area of fuel quality control, statistical control methods are now widely implemented replacing 100% inspection. This evolution, some practical examples and IAEA activities are described in the paper. The paper presents major findings of the latest IAEA Technical Meetings (TMs) and training courses in the area with emphasis on information received at the TM and training course held in 1999 and other latest publications to provide an overview of new developments in process/quality control, their implementation and results obtained including new approaches to QC

2000 Annual Progress Report for Fuels for Advanced CIDI Engines and Fuel Cells

Energy Technology Data Exchange (ETDEWEB)

Chalk, S.

2000-12-11

The Department of Energy's Office of Transportation Technologies Fiscal Year (FY) 2000 Annual Progress Report for the Fuels for Advanced CIDI Engines and Fuel Cells Program highlights progress achieved during FY 2000 and comprises 22 summaries of industry and National Laboratory projects that were conducted. The report provides an overview of the exciting work being conducted to tackle the tough technical challenges associated with developing clean burning fuels that will enable meeting the performance goals of the Emission Control R and D for Advanced CIDI Engines and the Transportation Fuel Cell Power Systems Programs. The summaries cover the effects of CIDI engine emissions and fuel cell power system performance, the effects of lubricants on engine emissions, the effects of fuel and consumed lubricants on exhaust emission control devices and the health and safety, materials compatibility, and economics of advanced petroleum-based fuels.

Advanced methods for fabrication of PHWR and LMFBR fuels

International Nuclear Information System (INIS)

Ganguly, C.

1988-01-01

For self-reliance in nuclear power, the Department of Atomic Energy (DAE), India is pursuing two specific reactor systems, namely the pressurised heavy water reactors (PHWR) and the liquid metal cooled fast breeder reactors (LMFBR). The reference fuel for PHWR is zircaloy-4 clad high density (≤ 96 per cent T.D.) natural UO 2 pellet-pins. The advanced PHWR fuels are UO 2 -PuO 2 (≤ 2 per cent), ThO 2 -PuO 2 (≤ 4 per cent) and ThO 2 -U 233 O 2 (≤ 2 per cent). Similarly, low density (≤ 85 per cent T.D.) (UPu)O 2 pellets clad in SS 316 or D9 is the reference fuel for the first generation of prototype and commercial LMFBRs all over the world. However, (UPu)C and (UPu)N are considered as advanced fuels for LMFBRs mainly because of their shorter doubling time. The conventional method of fabrication of both high and low density oxide, carbide and nitride fuel pellets starting from UO 2 , PuO 2 and ThO 2 powders is 'powder metallurgy (P/M)'. The P/M route has, however, the disadvantage of generation and handling of fine powder particles of the fuel and the associated problem of 'radiotoxic dust hazard'. The present paper summarises the state-of-the-art of advanced methods of fabrication of oxide, carbide and nitride fuels and highlights the author's experience on sol-gel-microsphere-pelletisation (SGMP) route for preparation of these materials. The SGMP process uses sol gel derived, dust-free and free-flowing microspheres of oxides, carbide or nitride for direct pelletisation and sintering. Fuel pellets of both low and high density, excellent microhomogeneity and controlled 'open' or 'closed' porosity could be fabricated via the SGMP route. (author). 5 tables, 14 figs., 15 refs

Siemens advance PWR fuel assemblies (HTP) and cladding

International Nuclear Information System (INIS)

Stout, R. B.; Woods, K. N.

1997-01-01

This paper describes the key features of the Siemens HTP (High Thermal Performance) fuel design, the current in-reactor performance of this advanced fuel assembly design, and the advanced cladding types available

Development of demonstration facility design technology for advanced nuclear fuel cycle process

International Nuclear Information System (INIS)

Cho, Il Je; You, G. S.; Choung, W. M.; Lee, E. P.; Hong, D. H.; Lee, W. K.; Ku, J. H.; Moon, S. I.; Kwon, K. C.; Lee, K. I. and other

2012-04-01

PRIDE Facility, pyroprocess mock-up facility, is the first facility that is operated in inert atmosphere in the country. By using the facility, the functional requirements and validity of pyroprocess technology and facility related to the advanced fuel cycle can be verified with a low cost. Then, PRIDE will contribute to evaluate the technology viability, proliferation resistance and possibility of commercialization of the pyroprocess technology. It is essential to develop design technologies for the advanced nuclear fuel cycle demonstration facilities and complete the detailed design of PRIDE facility with capabilities of the stringent inert atmosphere control, fully remote operation which are necessary to develop the high-temperature molten salts technology. For these, it is necessary to design the essential equipment of large scale inert cell structure and the control system to maintain the inert atmosphere, and evaluate the safety. To construct the hot cell system which is appropriate for pyroprocess, some design technologies should be developed, which include safety evaluation for effective operation and maintenance, radiation safety analysis for hot cell, structural analysis, environmental evaluation, HVAC systems and electric equipment

ACSEPT: a new FP7-Euratom Collaborative Project in the field of partitioning processes for advanced fuel cycles

International Nuclear Information System (INIS)

Bourg, Stephane; Touron, Emmanuel; Caravaca, Concha; Ekberg, Christian; Gaubert, Emmanuel; Hill, Clement

2008-01-01

Actinide recycling by separation and transmutation is considered worldwide and particularly in several European countries as one of the most promising strategies to reduce the inventory of radioactive waste, thus contributing to the sustainability of nuclear energy. Consistently with potentially viable recycling strategies, the Collaborative Project ACSEPT will provide a structured research and development framework to develop chemical separation processes compatible with fuel fabrication techniques, with a view to their future demonstration at the pilot level. Two strategies are proposed for the recycling of the actinides issuing from various forms of future nuclear fuels: -) their homogeneous recycling in mixed fuels via a prior group separation of the actinides and -) their heterogeneous recycling in targets or core blankets via their selective separation from fission products. Two major technologies are considered to meet these challenges: hydrometallurgical processes and pyrochemical processes. A training and education programme will also be implemented to share the knowledge among communities and generations so as to maintain the nuclear expertise at the fore-front of Europe. The challenging objectives of ACSEPT will be addressed by a multi-disciplinary consortium composed of European universities, nuclear research bodies and major industrial players. This consortium will generate fundamental improvements for the future design of a potential Advanced Processing Pilot Unit

Advanced PWR fuel design concepts

International Nuclear Information System (INIS)

Andersor, C.K.; Harris, R.P.; Crump, M.W.; Fuhrman, N.

1987-01-01

For nearly 15 years, Combustion Engineering has provided pressurized water reactor fuel with the features most suppliers are now introducing in their advanced fuel designs. Zircaloy grids, removable upper end fittings, large fission gas plenum, high burnup, integral burnable poisons and sophisticated analytical methods are all features of C-E standard fuel which have been well proven by reactor performance. C-E's next generation fuel for pressurized water reactors features 24-month operating cycles, optimal lattice burnable poisons, increased resistance to common industry fuel rod failure mechanisms, and hardware and methodology for operating margin improvements. Application of these various improvements offer continued improvement in fuel cycle economics, plant operation and maintenance. (author)

Advanced research reactor fuel development

Energy Technology Data Exchange (ETDEWEB)

Kim, Chang Kyu; Pak, H. D.; Kim, K. H. [and others

2000-05-01

The fabrication technology of the U{sub 3}Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U{sub 3}Si{sub 2} dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U{sub 3}Si{sub 2} fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 {approx} 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The {gamma}-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U{sub 3}Si{sub 2}. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano

Nuclear heat source design for an advanced HTGR process heat plant

International Nuclear Information System (INIS)

McDonald, C.F.; O'Hanlon, T.W.

1983-01-01

A high-temperature gas-cooled reactor (HTGR) coupled with a chemical process facility could produce synthetic fuels (i.e., oil, gasoline, aviation fuel, methanol, hydrogen, etc.) in the long term using low-grade carbon sources (e.g., coal, oil shale, etc.). The ultimate high-temperature capability of an advanced HTGR variant is being studied for nuclear process heat. This paper discusses a process heat plant with a 2240-MW(t) nuclear heat source, a reactor outlet temperature of 950 0 C, and a direct reforming process. The nuclear heat source outputs principally hydrogen-rich synthesis gas that can be used as a feedstock for synthetic fuel production. This paper emphasizes the design of the nuclear heat source and discusses the major components and a deployment strategy to realize an advanced HTGR process heat plant concept

Fuel processing

International Nuclear Information System (INIS)

Allardice, R.H.

1990-01-01

The technical and economic viability of the fast breeder reactor as an electricity generating system depends not only upon the reactor performance but also on a capability to recycle plutonium efficiently, reliably and economically through the reactor and fuel cycle facilities. Thus the fuel cycle is an integral and essential part of the system. Fuel cycle research and development has focused on demonstrating that the challenging technical requirements of processing plutonium fuel could be met and that the sometimes conflicting requirements of the fuel developer, fuel fabricator and fuel reprocessor could be reconciled. Pilot plant operation and development and design studies have established both the technical and economic feasibility of the fuel cycle but scope for further improvement exists through process intensification and flowsheet optimization. These objectives and the increasing processing demands made by the continuing improvement to fuel design and irradiation performance provide an incentive for continuing fuel cycle development work. (author)

Fast Reactor Fuel Cycle Cost Estimates for Advanced Fuel Cycle Studies

International Nuclear Information System (INIS)

Harrison, Thomas

2013-01-01

Presentation Outline: • Why Do I Need a Cost Basis?; • History of the Advanced Fuel Cycle Cost Basis; • Description of the Cost Basis; • Current Work; • Fast Reactor Fuel Cycle Applications; • Sample Fuel Cycle Cost Estimate Analysis; • Future Work

Advanced thermally stable jet fuels

Energy Technology Data Exchange (ETDEWEB)

Schobert, H.H.

1999-01-31

The Pennsylvania State University program in advanced thermally stable coal-based jet fuels has five broad objectives: (1) Development of mechanisms of degradation and solids formation; (2) Quantitative measurement of growth of sub-micrometer and micrometer-sized particles suspended in fuels during thermal stressing; (3) Characterization of carbonaceous deposits by various instrumental and microscopic methods; (4) Elucidation of the role of additives in retarding the formation of carbonaceous solids; (5) Assessment of the potential of production of high yields of cycloalkanes by direct liquefaction of coal. Future high-Mach aircraft will place severe thermal demands on jet fuels, requiring the development of novel, hybrid fuel mixtures capable of withstanding temperatures in the range of 400--500 C. In the new aircraft, jet fuel will serve as both an energy source and a heat sink for cooling the airframe, engine, and system components. The ultimate development of such advanced fuels requires a thorough understanding of the thermal decomposition behavior of jet fuels under supercritical conditions. Considering that jet fuels consist of hundreds of compounds, this task must begin with a study of the thermal degradation behavior of select model compounds under supercritical conditions. The research performed by The Pennsylvania State University was focused on five major tasks that reflect the objectives stated above: Task 1: Investigation of the Quantitative Degradation of Fuels; Task 2: Investigation of Incipient Deposition; Task 3: Characterization of Solid Gums, Sediments, and Carbonaceous Deposits; Task 4: Coal-Based Fuel Stabilization Studies; and Task 5: Exploratory Studies on the Direct Conversion of Coal to High Quality Jet Fuels. The major findings of each of these tasks are presented in this executive summary. A description of the sub-tasks performed under each of these tasks and the findings of those studies are provided in the remainder of this volume

Advanced coal-fueled gas turbine systems

Energy Technology Data Exchange (ETDEWEB)

Wenglarz, R.A.

1994-08-01

Several technology advances since the early coal-fueled turbine programs that address technical issues of coal as a turbine fuel have been developed in the early 1980s: Coal-water suspensions as fuel form, improved methods for removing ash and contaminants from coal, staged combustion for reducing NO{sub x} emissions from fuel-bound nitrogen, and greater understanding of deposition/erosion/corrosion and their control. Several Advanced Coal-Fueled Gas Turbine Systems programs were awarded to gas turbine manufacturers for for components development and proof of concept tests; one of these was Allison. Tests were conducted in a subscale coal combustion facility and a full-scale facility operating a coal combustor sized to the Allison Model 501-K industrial turbine. A rich-quench-lean (RQL), low nitrogen oxide combustor design incorporating hot gas cleanup was developed for coal fuels; this should also be applicable to biomass, etc. The combustor tests showed NO{sub x} and CO emissions {le} levels for turbines operating with natural gas. Water washing of vanes from the turbine removed the deposits. Systems and economic evaluations identified two possible applications for RQL turbines: Cogeneration plants based on Allison 501-K turbine (output 3.7 MW(e), 23,000 lbs/hr steam) and combined cycle power plants based on 50 MW or larger gas turbines. Coal-fueled cogeneration plant configurations were defined and evaluated for site specific factors. A coal-fueled turbine combined cycle plant design was identified which is simple, compact, and results in lower capital cost, with comparable efficiency and low emissions relative to other coal technologies (gasification, advanced PFBC).

Development of Demonstration Facility Design Technology for Advanced Nuclear Fuel Cycle Process

International Nuclear Information System (INIS)

Cho, Il Je; You, G. S.; Choung, W. M.

2010-04-01

The main objective of this R and D is to develop the PRIDE (PyRoprocess Integrated inactive DEmonstration) facility for engineering-scale inactive test using fresh uranium, and to establish the design requirements of the ESPF (Engineering Scale Pyroprocess Facility) for active demonstration of the pyroprocess. Pyroprocess technology, which is applicable to GEN-IV systems as one of the fuel cycle options, is a solution of the spent fuel accumulation problems. PRIDE Facility, pyroprocess mock-up facility, is the first facility that is operated in inert atmosphere in the country. By using the facility, the functional requirements and validity of pyroprocess technology and facility related to the advanced fuel cycle can be verified with a low cost. Then, PRIDE will contribute to evaluate the technology viability, proliferation resistance and possibility of commercialization of the pyroprocess technology. The PRIDE evaluation data, such as performance evaluation data of equipment and operation experiences, will be directly utilized for the design of ESPF

the effect of advanced fuel designs on fuel utilization

International Nuclear Information System (INIS)

Sarikaya, B.; Colak, U.; Tombakoglu, M.; Yilmazbayhan, A.

1997-01-01

Fuel management is one of the key topic in nuclear engineering. It is possible to increase fuel burnup and reactor lifetime by using advanced fuel management strategies. In order to increase the cycle lifetime, required amount of excess reactivity must be added to system. Burnable poisons can be used to compensate this excess reactivity. Usually gadolinium (Gd) is used as burnable poison. But the use of Gd presents some difficulties that have not been encountered with the use of boron

Development of Experimental Facilities for Advanced Spent Fuel Management Technology

Energy Technology Data Exchange (ETDEWEB)

You, G. S.; Jung, W. M.; Ku, J. H. [and others

2004-07-01

The advanced spent fuel management process(ACP), proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel, is under research and development. This technology convert spent fuels into pure metal-base uranium with removing the highly heat generating materials(Cs, Sr) efficiently and reducing of the decay heat, volume, and radioactivity from spent fuel by 1/4. In the next phase(2004{approx}2006), the demonstration of this technology will be carried out for verification of the ACP in a laboratory scale. For this demonstration, the hot cell facilities of {alpha}-{gamma} type and auxiliary facilities are required essentially for safe handling of high radioactive materials. As the hot cell facilities for demonstration of the ACP, a existing hot cell of {beta}-{gamma} type will be refurbished to minimize construction expenditures of hot cell facility. In this study, the design requirements are established, and the process detail work flow was analysed for the optimum arrangement to ensure effective process operation in hot cell. And also, the basic and detail design of hot cell facility and process, and safety analysis was performed to secure conservative safety of hot cell facility and process.

Study of advanced fuel system concepts for commercial aircraft

Science.gov (United States)

Coffinberry, G. A.

1985-01-01

An analytical study was performed in order to assess relative performance and economic factors involved with alternative advanced fuel systems for future commercial aircraft operating with broadened property fuels. The DC-10-30 wide-body tri-jet aircraft and the CF6-8OX engine were used as a baseline design for the study. Three advanced systems were considered and were specifically aimed at addressing freezing point, thermal stability and lubricity fuel properties. Actual DC-10-30 routes and flight profiles were simulated by computer modeling and resulted in prediction of aircraft and engine fuel system temperatures during a nominal flight and during statistical one-day-per-year cold and hot flights. Emergency conditions were also evaluated. Fuel consumption and weight and power extraction results were obtained. An economic analysis was performed for new aircraft and systems. Advanced system means for fuel tank heating included fuel recirculation loops using engine lube heat and generator heat. Environmental control system bleed air heat was used for tank heating in a water recirculation loop. The results showed that fundamentally all of the three advanced systems are feasible but vary in their degree of compatibility with broadened-property fuel.

Fuel behavior in advanced water reactors

International Nuclear Information System (INIS)

Bolme, A.B.

1996-01-01

Fuel rod behavior of advanced pressurized water reactors under steady state conditions has been investigated in this study. System-80+ and Westinghouse Vantage-5 fuels have been considered as advanced pressurized water reactor fuels to be analyzed. The purpose of this study is to analyze the sensitivity of ditferent models and the effect of selected design parameters on the overall fuel behavior. FRAPCON-II computer code has been used for the analyses. Different modelling options of FRAPCON-II have also been considered in these analyses. Analyses have been performed in two main parts. In the first part, effects of operating conditions on fuel behavior have been investigated. First, fuel rod response under normal operating conditions has been analyzed. Then, fuel rod response to different fuel ratings has been calculated. In the second part, in order to estimate the effect of design parameters on fuel behavior, parametric analyses have been performed. In this part, the effects of initial gap thickness, as fabricated fuel density, and initial fill gas pressure on fuel behavior have been analyzed. The computations showed that both of the fuel rods used in this study operate within the safety limits. However, FRAPCON-II modelling options have been resulted in different behavior due to their modelling characteristics. Hence, with the absence of experimental data, it is difficult to make assesment for the best fuel parameters. It is also difficult to estimate error associated with the results. To improve the performance of the code, it is necessary to develop better experimental correlations for material properties in order to analyze the eftect ot considerably different design parameters rather than nominal rod parameters

Recent advances in fuel fabrication techniques and prospects for the nineties

International Nuclear Information System (INIS)

Frain, R.G.; Caudill, H.L.; Faulhaber, R.

1987-01-01

Advanced Nuclear Fuels Corporation's approach and experience with the application of a flexible, just-in-time manufacturing philosophy to the production of customized nuclear fuel is described. Automation approaches to improve productivity are described. The transfer of technology across product lines is discussed as well as the challenges presented by a multiple product fabrication facility which produces a wide variety of BWR and PWR designs. This paper also describes the method of managing vendor quality control programs in support of standardization and clarity of documentation. Process simplification and the ensuing experience are discussed. Prospects for fabrication process advancements in the nineties are given with emphasis on the benefits of dry conversion of UF 6 to UO 2 powder, and increased use of automated and computerized inspection techniques. (author)

Thermodynamics of Advanced Fuels - International Database Project

International Nuclear Information System (INIS)

Massara, Simone; Gueneau, Christine

2014-01-01

The Thermodynamics of Advanced Fuels - International Database (TAF-ID) Project was established in 2013 under the auspices of the NEA Nuclear Science Committee. The project was designed to make available a comprehensive, internationally recognised and quality-assured database of phase diagrams and thermodynamic properties of advanced nuclear fuels with a view to meeting specialised requirements for the development of advanced fuels for a future generation of nuclear reactors. Some of the specific technical objectives that this programme intends to achieve are to predict the solid, liquid and/or gas phases formed during fuel cladding chemical interactions under normal and accident conditions, to improve the control of the experimental conditions during the fabrication of fuel materials at high temperature, for example by predicting the vapour pressures of the elements (particularly of plutonium and the minor actinides) and to predict the evolution of the chemical composition of fuel under irradiation versus temperature and burn-up. This joint project, co-ordinated by the NEA, was established for an initial three-year period among nine organisations from six NEA member countries: Canada (AECL, RMCC, UOIT), France (CEA), Japan (JAEA, CRIEPI), the Netherlands (NRG), the Republic of Korea (KAERI) and the United States (US DOE). It is entirely funded by the nine signatories of the project. (authors)

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

International Nuclear Information System (INIS)

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

2008-01-01

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

An Automated Process for Generation of New Fuel Breakdown Mechanisms

National Research Council Canada - National Science Library

Violi, Angela

2006-01-01

.... It combines advanced computational techniques in a synergistic study of the critical processes in fuel decomposition at a level of detail that can help distinguish, correct, and quantify mechanisms for these processes...

Treatment of waste salt from the advanced spent fuel conditioning process (II) : optimum immobilization condition

International Nuclear Information System (INIS)

Kim, Jeong Guk; Lee, Jae Hee; Yoo, Jae Hyung; Kim, Joon Hyung

2004-01-01

Since zeolite is known to be stable at a high temperature, it has been reported as a promising immobilization matrix for waste salt. The crystal structure of dehydrated zeolite A breaks down above 1060 K, resulting in the formation of an amorphous solid and re-crystallization to beta-Cristobalite. This structural degradation depends on the existence of chlorides. When contacted to HCl, zeolite 4A is not stable even at 473 K. The optimum consolidation condition for LiCl salt waste from the oxide fuel reduction process based on the electrochemical method (Advanced spent fuel Conditioning Process; ACP) has been studied using zeolite A since 2001. Actually the constituents of waste salt are water-soluble. And, alkali halides are known to be readily radiolyzed to yield interstitial halogens and metal colloids. For disposal in a geological repository, the waste salt must meet the acceptance criteria. For a waste form containing chloride salt, two of the more important criteria are leach resistance and waste form durability. In this work, we prepared some samples with different mixing ratios of LiCl salt to zeolite A, and then compared some characteristics such as thermal stability, salt occlusion, free chloride content, leach resistance, mixing effect, etc

Cermet-fueled reactors for advanced space applications

International Nuclear Information System (INIS)

Cowan, C.L.; Palmer, R.S.; Taylor, I.N.; Vaidyanathan, S.; Bhattacharyya, S.K.; Barner, J.O.

1987-12-01

Cermet-fueled nuclear reactors are attractive candidates for high-performance advanced space power systems. The cermet consists of a hexagonal matrix of a refractory metal and a ceramic fuel, with multiple tubular flow channels. The high performance characteristics of the fuel matrix come from its high strength at elevated temperatures and its high thermal conductivity. The cermet fuel concept evolved in the 1960s with the objective of developing a reactor design that could be used for a wide range of mobile power generating sytems, including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests for the cermet fuel were carried out by General Electric as part of the 710 Project (General Electric 1966), and by Argonne National Laboratory in the Direct Nuclear Rocket Program (1965). Development programs for cermet fuel are currently under way at Argonne National Laboratory and Pacific Northwest Laboratory. The high temperature qualification tests from the 1960s have provided a base for the incorporation of cermet fuel in advanced space applications. The status of the cermet fuel development activities and descriptions of the key features of the cermet-fueled reactor design are summarized in this paper

Determination of Basic Structure-Property Relations for Processing and Modeling in Advanced Nuclear Fuel: Microstructure Evolution and Mechanical Properties

International Nuclear Information System (INIS)

Wheeler, Kirk; Parra, Manuel; Peralta, Pedro

2009-01-01

The project objective is to study structure-property relations in solid solutions of nitrides and oxides with surrogate elements to simulate the behavior of fuels of inert matrix fuels of interest to the Advanced Fuel Cycle Initiative (AFCI), with emphasis in zirconium-based materials. Work with actual fuels will be carried out in parallel in collaboration with Los Alamos National Laboratory (LANL). Three key aspects will be explored: microstructure characterization through measurement of global texture evolution and local crystallographic variations using Electron Backscattering Diffraction (EBSD); determination of mechanical properties, including fracture toughness, quasi-static compression strength, and hardness, as functions of load and temperature, and, finally, development of structure-property relations to describe mechanical behavior of the fuels based on experimental data. Materials tested will be characterized to identify the mechanisms of deformation and fracture and their relationship to microstructure and its evolution. New aspects of this research are the inclusion of crystallographic information into the evaluation of fuel performance and the incorporation of statistical variations of microstructural variables into simplified models of mechanical behavior of fuels that account explicitly for these variations. The work is expected to provide insight into processing conditions leading to better fuel performance and structural reliability during manufacturing and service, as well as providing a simplified testing model for future fuel production

Thermochemistry of nuclear fuels in advanced reactors

International Nuclear Information System (INIS)

Agarwal, Renu

2015-01-01

The presence of a large number of elements, accompanied with steep temperature gradient results in dynamic chemistry during nuclear fuel burn-up. Understanding this chemistry is very important for efficient and safe usage of nuclear fuels. The radioactive nature of these fuels puts lot of constraint on regulatory bodies to ensure their accident free operation in the reactors. One of the common aims of advanced fuels is to achieve high burn-up. As burn-up of the fuel increases, chemistry of fission-products becomes increasingly more important. To understand different phenomenon taking place in-pile, many out of-pile experiments are carried out. Extensive studies of thermodynamic properties, phase analysis, thermophysical property evaluation, fuel-fission product clad compatibility are carried out with relevant compounds and simulated fuels (SIMFUEL). All these data are compiled and jointly evaluated using different computational methods to predict fuel behaviour during burn-up. Only when this combined experimental and theoretical information confirms safe operation of the pin, a test pin is prepared and burnt in a test reactor. Every fuel has a different chemistry and different constraints associated with it. In this talk, various thermo-chemical aspects of some of the advanced fuels, mixed carbide, mixed nitride, 'Pu' rich MOX, 'Th' based AHWR fuels and metallic fuels will be discussed. (author)

Advanced Fuel Cycle Economic Sensitivity Analysis

Energy Technology Data Exchange (ETDEWEB)

David Shropshire; Kent Williams; J.D. Smith; Brent Boore

2006-12-01

A fuel cycle economic analysis was performed on four fuel cycles to provide a baseline for initial cost comparison using the Gen IV Economic Modeling Work Group G4 ECON spreadsheet model, Decision Programming Language software, the 2006 Advanced Fuel Cycle Cost Basis report, industry cost data, international papers, the nuclear power related cost study from MIT, Harvard, and the University of Chicago. The analysis developed and compared the fuel cycle cost component of the total cost of energy for a wide range of fuel cycles including: once through, thermal with fast recycle, continuous fast recycle, and thermal recycle.

Design study of advanced nuclear fuel recycle system. Conceptual study of recycle system using molten salt

International Nuclear Information System (INIS)

Kakehi, I.; Shirai, N.; Hatano, M.; Kajitani, M.; Yonezawa, S.; Kawai, T.; Kawamura, F.; Tobe, K.; Takahashi, K.

1996-12-01

For the purpose of developing the future nuclear fuel recycle system, the design study of the advanced nuclear fuel recycle system is being conducted. This report describes intermediate accomplishments in the conceptual system study of the advanced nuclear fuel recycle system. Fundamental concepts of this system is the recycle system using molten salt which intend to break through the conventional concepts of purex and pellet fuel system. Contents of studies in this period are as follows, 1)feasibility study of the process by Cd-cathode for nitride fuel, 2)application study for the molten salt of low melting point (AlCl3+organic salt), 3)research for decladding (advantage of decladding by heat treatment), 4)behavior of FPs in electrorefining (behavior of iodine and volatile FP chlorides, FPs behavior in chlorination), 5)criticality analysis in electrorefiner, 6)drawing of off-gas flow diagram, 7)drawing of process machinery concept (cathode processor, vibration packing), 8)evaluation for the amounts of the high level radioactive wastes, 9)quality of the recycle fuels (FPs contamination of recycle fuel), 10)conceptual study of in-cell handling system, 11)meaning of the advanced nuclear fuel recycle system. The conceptual system study will be completed in describing concepts of the system and discussing issues for the developments. (author)

Fuels processing for transportation fuel cell systems

Science.gov (United States)

Kumar, R.; Ahmed, S.

Fuel cells primarily use hydrogen as the fuel. This hydrogen must be produced from other fuels such as natural gas or methanol. The fuel processor requirements are affected by the fuel to be converted, the type of fuel cell to be supplied, and the fuel cell application. The conventional fuel processing technology has been reexamined to determine how it must be adapted for use in demanding applications such as transportation. The two major fuel conversion processes are steam reforming and partial oxidation reforming. The former is established practice for stationary applications; the latter offers certain advantages for mobile systems and is presently in various stages of development. This paper discusses these fuel processing technologies and the more recent developments for fuel cell systems used in transportation. The need for new materials in fuels processing, particularly in the area of reforming catalysis and hydrogen purification, is discussed.

Study of Advanced Reactor Mixed Oxide Fuel Production of (U,Th)O2

International Nuclear Information System (INIS)

Busron-Masduki; Damunir; Pristi-Hartati; R-Sukarsono; Bangun-Wasito

2000-01-01

The high price and starting scarcity of reserved of oil drive the people to drill the alternative nuclear energy. Accelerator-driven Transmutation Waste (ATW) is a prospective technology to solve the problem of used fuel waste, to reduce the anxiety of long term disposal waste, to increase the public acceptance of nuclear energy enter into the third millennium. The future of large nuclear energy appears in many-branched industry will depend on the capability to generate relatively low priced fuel on the basis of commercial nuclear energy. Utilization of uranium-233 -thorium cycle insures long-term fuel supply, makes the nuclear energy production more flexible and enables the self-provision regime to be realized in future. Flowsheet of mixed oxide fuel production for advanced reactor of (U,Th)O 2 is a combination of existing manufacturing equipment and quality assurance program from commercial LWR and HTR. The front-end of flowsheet using sol-gel process. The external sol-gel process is chosen due to simple equipment can anticipate refabrication of U-233 which always contains a few hundred ppm of U-232 and its gamma-emitting daughters, besides yielding smaller waste. The decision to choose external sol-gel process encourages to develop External Gelation Thorium (EGT). In order to get higher density and relatively low compaction pressures (i.e. for advanced LWR) adopted flowsheet EGT is developed to be Sol-Gel Microsphere Pelletization (SGMP). Using the optimal parameters, SGMP become established flowsheet for producing mixed oxide fuel of (U,Th)O 2 for advanced reactor. (author)

Advanced-fuel reversed-field pinch reactor (RFPR)

International Nuclear Information System (INIS)

Hagenson, R.L.; Krakowski, R.A.

1981-10-01

The utilization of deuterium-based fuels offers the potential advantages of greater flexibility in blanket design, significantly reduced tritium inventory, potential reduction in radioactivity level, and utilization of an inexhaustible fuel supply. The conventional DT-fueled Reversed-Field Pinch Reactor (RFPR) designs are reviewed, and the recent extension of these devices to advanced-fuel (catalyzed-DD) operation is presented. Attractive and economically competitive DD/RFPR systems are identified having power densities and plasma parameters comparable to the DT systems. Converting an RFP reactor from DT to DD primarily requires increasing the magnetic field levels a factor of two, still requiring only modest magnet coil fields (less than or equal to 4 T). When compared to the mainline tokamak, the unique advantages of the RFP (e.g., high beta, low fields at the coils, high ohmic-heating power densities, unrestricted aspect ratio) are particularly apparent for the utilization of advanced fuels

Advanced Fuel Cycle Cost Basis – 2017 Edition

Energy Technology Data Exchange (ETDEWEB)

Dixon, B. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ganda, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Williams, K. A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hanson, J. K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-09-29

This report, commissioned by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the DOE Nuclear Technology Research and Development (NTRD) Program (previously the Fuel Cycle Research and Development (FCRD) and the Advanced Fuel Cycle Initiative (AFCI)). The report describes the NTRD cost basis development process, reference information on NTRD 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 numerous fuel cycle cost modules (modules A-O) as well as cost modules for a number of reactor types (R modules). The fuel cycle cost modules were developed in the areas of natural uranium mining and milling, thorium 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, managed decay storage, recycled product 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. Since its inception, this report has been periodically updated. The last such internal document was published in August 2015 while the last external edition was published in December of 2009 as INL/EXT-07-12107 and is available on the Web at URL: www.inl.gov/technicalpublications/Documents/4536700.pdf. This current report (Sept 2017) is planned to be reviewed for external release, at which time it will replace the 2009 report as an external publication. This information is used in the ongoing evaluation of nuclear fuel cycles by the NE NTRD program.

Developing Spent Fuel Assembly for Advanced NDA Instrument Calibration - NGSI Spent Fuel Project

Energy Technology Data Exchange (ETDEWEB)

Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Banfield, James [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Skutnik, Steven [Univ. of Tennessee, Knoxville, TN (United States)

2014-02-01

This report summarizes the work by Oak Ridge National Laboratory to investigate the application of modeling and simulation to support the performance assessment and calibration of the advanced nondestructive assay (NDA) instruments developed under the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Advanced NDA instrument calibration will likely require reference spent fuel assemblies with well-characterized nuclide compositions that can serve as working standards. Because no reference spent fuel standard currently exists, and the practical ability to obtain direct measurement of nuclide compositions using destructive assay (DA) measurements of an entire fuel assembly is prohibitive in the near term due to the complexity and cost of spent fuel experiments, modeling and simulation will be required to construct such reference fuel assemblies. These calculations will be used to support instrument field tests at the Swedish Interim Storage Facility (Clab) for Spent Nuclear Fuel.

Economics of coal conversion processing. Advances in coal gasification: support research. Advances in coal gasification: process development and analysis

Energy Technology Data Exchange (ETDEWEB)

1978-01-01

The fall meeting of the American Chemical Society, Division of Fuel Chemistry, was held at Miami Beach, Florida, September 10-15, 1978. Papers involved the economics of coal conversion processing and advances in coal gasification, especially support research and process development and analysis. Fourteen papers have been entered individually into EDB and ERA; three papers had been entered previously from other sources. (LTN)

ACSEPT a European project for a new step in the future demonstration of advanced fuel processing

International Nuclear Information System (INIS)

Bourg, S.; Hill, C.; Caravaca, C.; Espartero, A.; Rhodes, C.; Taylor, R.; Harrison, M.; EKBERG, C.; GEIST, A.; Modolo, G.; Cassayre, L.; Malmbeck, R.; De Angelis, G.; Bouvet, S.; Klaassen, F.

2010-01-01

For more than fifteen years, a European scientific community has joined its effort to develop and optimise processes for the partitioning of actinides from fission products. In an international context of 'nuclear renaissance', the upcoming of a new generation of nuclear reactor (Gen IV) will require the development of associated advanced closed fuel cycles which answer the needs of a sustainable nuclear energy: the minimization of the production of long lived radioactive waste but also the optimization of the use of natural resources with an increased resistance to proliferation. Actually, Partitioning and Transmutation (P and T), associated to a multi-recycling of all transuranics (TRUs), should play a key role in the development of this sustainable nuclear energy. By joining together 34 Partners coming from European universities, nuclear research bodies and major industrial players in a multidisciplinary consortium, the FP7 EURATOM-Fission Collaborative Project ACSEPT (Actinide recycling by Separation and Transmutation), started in 2008 for four year duration, provides the sound basis and fundamental improvements for future demonstrations of fuel treatment in strong connection with fuel fabrication techniques. Consistently with potentially viable recycling strategies, ACSEPT therefore provides a structured R and D framework to develop chemical separation processes compatible with fuel fabrication techniques, with a view to their future demonstration at the pilot level. ACSEPT is organized into three technical domains: (i) Considering technically mature aqueous separation processes, ACSEPT works to optimize and select the most promising ones dedicated either to actinide partitioning or to group actinide separation. (ii) Concerning high temperature pyrochemical separation processes, ACSEPT focuses on the enhancement of the two reference cores of process selected within previous projects. R and D efforts are now devoted to key scientific and technical points

Chemical compatibility between cladding alloys and advanced fuels

International Nuclear Information System (INIS)

Fee, D.C.; Johnson, C.E.

1975-05-01

The National Advanced Fuels Program requires chemical, mechanical, and thermophysical properties data for cladding alloys. The compatibility behavior of cladding alloys with advanced fuels is critically reviewed. in carbide fuel pins, the principal compatibility problem is cladding carburization, diffusion of carbon into the cladding matrix accompanied by carbide precipitation. Carburization changes the mechanical properties of the cladding alloy. The extent of carburization increases in sodium (versus gas) bonded fuels. The depth of carburization increases with increasing sesquicarbide (M 2 C 3 ) content of the fuel. In nitride fuel pins, the principal compatibility problem is cladding nitriding, diffusion of nitrogen into the cladding matrix accompanied by nitride precipitation. Nitriding changes the mechanical properties of the cladding alloy. In both carbide and nitride fuel pins, fission products do not migrate appreciably to the cladding and do not appear to contribute to cladding attack. 77 references. (U.S.)

Recent advances during the treatment of spent EBR-II fuel

International Nuclear Information System (INIS)

Westphal, B.R.; Mariani, R.D.; Vaden, D.E.; Sherman, S.R.; Li, S.X.; Keiser, D.D. Jr.

2000-01-01

Several recent advances have been achieved for the electrometallurgical treatment of spent nuclear fuel. In anticipation of production operations at Argonne National Laboratory-West, development of both electrorefining and metal processing has been ongoing in the post-demonstration phase in order to further optimize the process. These development activities show considerable promise. This paper discusses the results of recent experiments as well as plans for future investigations

Advanced Fuels Campaign Execution Plan

Energy Technology Data Exchange (ETDEWEB)

Kemal Pasamehmetoglu

2010-10-01

The purpose of the Advanced Fuels Campaign (AFC) Execution Plan is to communicate the structure and management of research, development, and demonstration (RD&D) activities within the Fuel Cycle Research and Development (FCRD) program. Included in this document is an overview of the FCRD program, a description of the difference between revolutionary and evolutionary approaches to nuclear fuel development, the meaning of science-based development of nuclear fuels, and the “Grand Challenge” for the AFC that would, if achieved, provide a transformational technology to the nuclear industry in the form of a high performance, high reliability nuclear fuel system. The activities that will be conducted by the AFC to achieve success towards this grand challenge are described and the goals and milestones over the next 20 to 40 year period of research and development are established.

Advanced Fuels Campaign Execution Plan

Energy Technology Data Exchange (ETDEWEB)

Kemal Pasamehmetoglu

2011-09-01

The purpose of the Advanced Fuels Campaign (AFC) Execution Plan is to communicate the structure and management of research, development, and demonstration (RD&D) activities within the Fuel Cycle Research and Development (FCRD) program. Included in this document is an overview of the FCRD program, a description of the difference between revolutionary and evolutionary approaches to nuclear fuel development, the meaning of science-based development of nuclear fuels, and the 'Grand Challenge' for the AFC that would, if achieved, provide a transformational technology to the nuclear industry in the form of a high performance, high reliability nuclear fuel system. The activities that will be conducted by the AFC to achieve success towards this grand challenge are described and the goals and milestones over the next 20 to 40 year period of research and development are established.

Fuel rod bundles proposed for advanced pressure tube nuclear reactors

International Nuclear Information System (INIS)

Prodea, Iosif; Catana, Alexandru

2010-01-01

The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACR TM -1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)

Lessons Learned From Dynamic Simulations of Advanced Fuel Cycles

International Nuclear Information System (INIS)

Piet, Steven J.; Dixon, Brent W.; Jacobson, Jacob J.; Matthern, Gretchen E.; Shropshire, David E.

2009-01-01

Years of performing dynamic simulations of advanced nuclear fuel cycle options provide insights into how they could work and how one might transition from the current once-through fuel cycle. This paper summarizes those insights from the context of the 2005 objectives and goals of the Advanced Fuel Cycle Initiative (AFCI). Our intent is not to compare options, assess options versus those objectives and goals, nor recommend changes to those objectives and goals. Rather, we organize what we have learned from dynamic simulations in the context of the AFCI objectives for waste management, proliferation resistance, uranium utilization, and economics. Thus, we do not merely describe 'lessons learned' from dynamic simulations but attempt to answer the 'so what' question by using this context. The analyses have been performed using the Verifiable Fuel Cycle Simulation of Nuclear Fuel Cycle Dynamics (VISION). We observe that the 2005 objectives and goals do not address many of the inherently dynamic discriminators among advanced fuel cycle options and transitions thereof

Advanced fuel technology and performance: Current status and trends

International Nuclear Information System (INIS)

1990-11-01

During the last years the Nuclear Fuel Cycle and Waste Management Division of the IAEA has been giving great attention to the collection, analysis and exchange of information in the field of reactor fuel technology. Most of these activities are being conducted in the framework of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT). The purpose of this Advisory Group Meeting on Advanced Fuel Technology and Performance was to update and to continue the previous work, and to review the experience of advanced fuel technology, its performance with regard to all types of reactors and to outline the future trends on the basis of national experience and discussions during the meeting. As a result of the meeting a Summary Report was prepared which reflected the status of the advanced nuclear fuel technology up to 1990. The 10 papers presented by participants of this meeting are also published here. A separate abstract was prepared for each of these papers. Refs, figs and tabs

Verification tests for CANDU advanced fuel

International Nuclear Information System (INIS)

Chung, Chang Hwan; Chang, S.K.; Hong, S.D.

1997-07-01

For the development of a CANDU advanced fuel, the CANFLEX-NU fuel bundles were tested under reactor operating conditions at the CANDU-Hot test loop. This report describes test results and test methods in the performance verification tests for the CANFLEX-NU bundle design. The main items described in the report are as follows. - Fuel bundle cross-flow test - Endurance fretting/vibration test - Freon CHF test - Production of technical document. (author). 25 refs., 45 tabs., 46 figs

Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study

Energy Technology Data Exchange (ETDEWEB)

Kristine Barrett; Shannon Bragg-Sitton

2012-09-01

The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system that would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.

Research investigations in oil shale, tar sand, coal research, advanced exploratory process technology, and advanced fuels research: Volume 1 -- Base program. Final report, October 1986--September 1993

Energy Technology Data Exchange (ETDEWEB)

Smith, V.E.

1994-05-01

Numerous studies have been conducted in five principal areas: oil shale, tar sand, underground coal gasification, advanced process technology, and advanced fuels research. In subsequent years, underground coal gasification was broadened to be coal research, under which several research activities were conducted that related to coal processing. The most significant change occurred in 1989 when the agreement was redefined as a Base Program and a Jointly Sponsored Research Program (JSRP). Investigations were conducted under the Base Program to determine the physical and chemical properties of materials suitable for conversion to liquid and gaseous fuels, to test and evaluate processes and innovative concepts for such conversions, to monitor and determine environmental impacts related to development of commercial-sized operations, and to evaluate methods for mitigation of potential environmental impacts. This report is divided into two volumes: Volume 1 consists of 28 summaries that describe the principal research efforts conducted under the Base Program in five topic areas. Volume 2 describes tasks performed within the JSRP. Research conducted under this agreement has resulted in technology transfer of a variety of energy-related research information. A listing of related publications and presentations is given at the end of each research topic summary. More specific and detailed information is provided in the topical reports referenced in the related publications listings.

Strategies in development of advanced fuels for LMFBR

International Nuclear Information System (INIS)

Handa, Muneo

1976-12-01

Overseas strategies in development of advanced fuels for LMFBR are reviewed. Recent irradiation experiment and out-of-pile test data of the fuels are given in detail. The present status of development of oxide fueled LMFBR is also treated. (auth.)

Advanced materials for alternative fuel capable directly fired heat engines

Energy Technology Data Exchange (ETDEWEB)

Fairbanks, J.W.; Stringer, J. (eds.)

1979-12-01

The first conference on advanced materials for alternative fuel capable directly fired heat engines was held at the Maine Maritime Academy, Castine, Maine. It was sponsored by the US Department of Energy, (Assistant Secretary for Fossil Energy) and the Electric Power Research Institute, (Division of Fossil Fuel and Advanced Systems). Forty-four papers from the proceedings have been entered into EDB and ERA and one also into EAPA; three had been entered previously from other sources. The papers are concerned with US DOE research programs in this area, coal gasification, coal liquefaction, gas turbines, fluidized-bed combustion and the materials used in these processes or equipments. The materials papers involve alloys, ceramics, coatings, cladding, etc., and the fabrication and materials listing of such materials and studies involving corrosion, erosion, deposition, etc. (LTN)

A Post Closure Safety Assessment for Radioactive Wastes from Advanced nuclear fuel Cycle

International Nuclear Information System (INIS)

Kang, Chul Hyung; Hwang, Yong Soo

2010-01-01

KAERI has developed the KIEP-21 (Korean, Innovative, Environmentally Friendly, and Proliferation Resistant System for the 21st Century). It is an advanced nuclear fuel cycle option with a pyro-process and a GEN-IV SFR. A pyro-process consists of two distinctive processes, an electrolytic reduction process and an electro-refining and winning process. When the pyro-process is applied, it generates five streams of wastes. To compare pyro-process advantage over the direct disposal of Spent Nuclear Fuel (SNF), the PWR SNF of the 45,000 MWD burn-up has been assumed. A safety assessment model for pyro-process wastes and representative results are presented in this report

Advanced disassembling technique of irradiated driver fuel assembly for continuous irradiation of fuel pins

International Nuclear Information System (INIS)

Ichikawa, Shoichi; Haga, Hiroyuki; Katsuyama, Kozo; Maeda, Koji; Nishinoiri, Kenji

2012-01-01

It was necessary to carry out continuous irradiation tests in order to obtain the irradiation data of high burn-up fuel and high neutron dose material for FaCT (Fast Reactor Cycle Technology Development) project. There, the disassembling technique of an irradiated fuel assembly was advanced in order to realize further continuous irradiation tests. Although the conventional disassembling technique had been cutting a lower end-plug of a fuel pin needed to fix fuel pins to an irradiation vehicle, the advanced disassembling technique did not need cutting a lower end-plug. As a result, it was possible to supply many irradiated fuel pins to various continuous irradiation tests for FaCT project. (author)

Advanced compressed hydrogen fuel storage systems

International Nuclear Information System (INIS)

Jeary, B.

2000-01-01

Dynetek was established in 1991 by a group of private investors, and since that time efforts have been focused on designing, improving, manufacturing and marketing advanced compressed fuel storage systems. The primary market for Dynetek fuel systems has been Natural Gas, however as the automotive industry investigates the possibility of using hydrogen as the fuel source solution in Alternative Energy Vehicles, there is a growing demand for hydrogen storage on -board. Dynetek is striving to meet the needs of the industry, by working towards developing a fuel storage system that will be efficient, economical, lightweight and eventually capable of storing enough hydrogen to match the driving range of the current gasoline fueled vehicles

Technical verification of advanced nuclear fuel for KSNPs

International Nuclear Information System (INIS)

Lee, C. B.; Bang, J. G.; Kim, D. H. and others

2002-03-01

KNFC has developed the advanced 16x16 fuel assembly for the Korean Standard Nuclear Plants through the three-year R and D project (from April 1999 to March 2002) under the Nuclear R and D program by MOST. The purpose of this project is to verify the advanced 16x16 fuel assembly for the Korean Standard Nuclear Plants being developed by KNFC during the same period. Verification tests for the advanced fuel assembly and its components such as characteristic test on the spacer grid spring and dimple, static buckling and dynamic impact test on the 5x5 partial spacer grid, the fuel rod vibration test supported by the PLUS7 mid-spacer grid, fretting wear test, turbulent flow structure test in wind tunnel and corrosion test were performed by using the KAERI facilities. Design reports and test results produced by KNFC were technically reviewed. For the domestic production of burnable poison rod, manufacturing technology of burnable poison pellets was developed

Analysis of the ATR fuel element swaging process

International Nuclear Information System (INIS)

Richins, W.D.; Miller, G.K.

1995-12-01

This report documents a detailed evaluation of the swaging process used to connect fuel plates to side plates in Advanced Test Reactor (ATR) fuel elements. The swaging is a mechanical process that begins with fitting a fuel plate into grooves in the side plates. Once a fuel plate is positioned, a lip on each of two side plate grooves is pressed into the fuel plate using swaging wheels to form the joints. Each connection must have a specified strength (measured in terms, of a pullout force capacity) to assure that these joints do not fail during reactor operation. The purpose of this study is to analyze the swaging process and associated procedural controls, and to provide recommendations to assure that the manufacturing process produces swaged connections that meet the minimum strength requirement. The current fuel element manufacturer, Babcock and Wilcox (B ampersand W) of Lynchburg, Virginia, follows established procedures that include quality inspections and process controls in swaging these connections. The procedures have been approved by Lockheed Martin Idaho Technologies and are designed to assure repeatability of the process and structural integrity of each joint. Prior to July 1994, ATR fuel elements were placed in the Hydraulic Test Facility (HTF) at the Idaho National Engineering Laboratory (AGNAIL), Test Reactor Area (TRA) for application of Boehmite (an aluminum oxide) film and for checking structural integrity before placement of the elements into the ATR. The results presented in this report demonstrate that the pullout strength of the swaged connections is assured by the current manufacturing process (with several recommended enhancements) without the need for- testing each element in the HTF

Advancing PWR fuel to meet customer needs

Energy Technology Data Exchange (ETDEWEB)

Kramer, F W

1987-03-01

Since the introduction of the Optimized Fuel Assembly (OFA) for PWRs in the late 1970s, Westinghouse has continued to work with the utility customers to identify the greatest needs for further advance in fuel performance and reliability. The major customer requirements include longer fuel cycle at lower costs, increased fuel discharge burn-up, enhanced operating flexibility, all accompanied by even greater reliability. In response to these needs, Westinghouse developed Vantage 5 PWR fuel. To optimize reactor operations, Vantage 5 fuel features distinct advantages: integral fuel burnable absorbers, axial and radial blankets, intermediate flow mixers, a removable top nozzle, and assembly modifications to accommodate increased discharge burn-up.

Advanced Combustion and Fuels; NREL (National Renewable Energy Laboratory)

Energy Technology Data Exchange (ETDEWEB)

Zigler, Brad

2015-06-08

Presented at the U.S. Department of Energy Vehicle Technologies Office 2015 Annual Merit Review and Peer Evaluation Meeting, held June 8-12, 2015, in Arlington, Virginia. It addresses technical barriers of inadequate data and predictive tools for fuel and lubricant effects on advanced combustion engines, with the strategy being through collaboration, develop techniques, tools, and data to quantify critical fuel physico-chemical effects to enable development of advanced combustion engines that use alternative fuels.

Advanced ceramic cladding for water reactor fuel

International Nuclear Information System (INIS)

Feinroth, H.

2000-01-01

Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

Treatment of waste salt from the advanced spent fuel conditioning process (I): characterization of Zeolite A in Molten LiCl Salt

International Nuclear Information System (INIS)

Kim, Jeong Guk; Lee, Jae Hee; Yoo, Jae Hyung; Kim, Joon Hyung

2004-01-01

The oxide fuel reduction process based on the electrochemical method (Advanced spent fuel Conditioning Process; ACP) and the long-lived radioactive nuclides partitioning process based on electro-refining process, which are being developed ay the Korea Atomic Energy Research Institute (KAERI), are to generate two types of molten salt wastes such as LiCl salt and LiCl-KCl eutectic salt, respectively. These waste salts must meet some criteria for disposal. A conditioning process for LiCl salt waste from ACP has been developed using zeolite A. This treatment process of waste salt using zeolite A was first developed by US ANL (Argonne National Laboratory) for LiCl-KCl eutectic salt waste from an electro-refining process of EBR (Experimental Breeder Reactor)-II spent fuel. This process has been developed recently, and a ceramic waste form (CWF) is produced in demonstration-scale V-mixer (50 kg/batch). However, ANL process is different from KAERI treatment process in waste salt, the former is LiCl-KCl eutectic salt and the latter is LiCl salt. Because of melting point, the immobilization of eutectic salt is carried out at about 770 K, whereas LiCl salt at around 920 K. Such difference has an effect on properties of immobilization media, zeolite A. Here, zeolite A in high-temperature (923 K) molten LiCl salt was characterized by XRD, Ion-exchange, etc., and evaluated if a promising media or not

Hydrodeoxygenation processes: advances on catalytic transformations of biomass-derived platform chemicals into hydrocarbon fuels.

Science.gov (United States)

De, Sudipta; Saha, Basudeb; Luque, Rafael

2015-02-01

Lignocellulosic biomass provides an attractive source of renewable carbon that can be sustainably converted into chemicals and fuels. Hydrodeoxygenation (HDO) processes have recently received considerable attention to upgrade biomass-derived feedstocks into liquid transportation fuels. The selection and design of HDO catalysts plays an important role to determine the success of the process. This review has been aimed to emphasize recent developments on HDO catalysts in effective transformations of biomass-derived platform molecules into hydrocarbon fuels with reduced oxygen content and improved H/C ratios. Liquid hydrocarbon fuels can be obtained by combining oxygen removal processes (e.g. dehydration, hydrogenation, hydrogenolysis, decarbonylation etc.) as well as by increasing the molecular weight via C-C coupling reactions (e.g. aldol condensation, ketonization, oligomerization, hydroxyalkylation etc.). Fundamentals and mechanistic aspects of the use of HDO catalysts in deoxygenation reactions will also be discussed. Copyright © 2014 Elsevier Ltd. All rights reserved.

Engineering development of advanced physical fine coal cleaning for premium fuel applications

International Nuclear Information System (INIS)

1997-01-01

Bechtel, together with Amax Research and Development Center (Amax R ampersand D), has prepared this study which provides conceptual cost estimates for the production of premium quality coal-water slurry fuel (CWF) in a commercial plant. Two scenarios are presented, one using column flotation technology and the other the selective agglomeration to clean the coal to the required quality specifications. This study forms part of US Department of Energy program Engineering Development of Advanced Physical Fine Coal Cleaning for Premium Fuel Applications, (Contract No. DE-AC22- 92PC92208), under Task 11, Project Final Report. The primary objective of the Department of Energy program is to develop the design base for prototype commercial advanced fine coal cleaning facilities capable of producing ultra-clean coals suitable for conversion to stable and highly loaded CWF. The fuels should contain less than 2 lb ash/MBtu (860 grams ash/GJ) of HHV and preferably less than 1 lb ash/MBtu (430 grams ash/GJ). The advanced fine coal cleaning technologies to be employed are advanced column froth flotation and selective agglomeration. It is further stipulated that operating conditions during the advanced cleaning process should recover not less than 80 percent of the carbon content (heating value) in the run-of-mine source coal. These goals for ultra-clean coal quality are to be met under the constraint that annualized coal production costs does not exceed $2.5 /MBtu ($ 2.37/GJ), including the mine mouth cost of the raw coal. A further objective of the program is to determine the distribution of a selected suite of eleven toxic trace elements between product CWF and the refuse stream of the cleaning processes. Laboratory, bench-scale and Process Development Unit (PDU) tests to evaluate advanced column flotation and selective agglomeration were completed earlier under this program with selected coal samples. A PDU with a capacity of 2 st/h was designed by Bechtel and installed at

IEA-Advanced Motor Fuels Annual Report 2012

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-06-15

The annual report from the IEA implementing agreement on Advanced Motor Fuels (AMF) describes what the agreement is about, how to join, various activities of the agreement, a message from the Chairman, and projects/annexes active for the year. An annual section covers the global situation for the topic of advanced motor fuels. Another section includes highlights coming from each country participating in AMF, and major sections relaying activities on each of the ongoing annexes. Information regarding participating delegations, contact information, publications resulting from AMF, and upcoming meetings rounds out the report.

IEA-Advanced Motor Fuels Annual Report 2011

Energy Technology Data Exchange (ETDEWEB)

NONE

2011-07-01

The annual report from the IEA implementing agreement on Advanced Motor Fuels (AMF) describes what the agreement is about, how to join, various activities of the agreement, a message from the Chairman, and projects/annexes active for the year. An annual section covers the global situation for the topic of advanced motor fuels. Another section includes highlights coming from each country participating in AMF, and major sections relaying activities on each of the ongoing annexes. Information regarding participating delegations, contact information, publications resulting from AMF, and upcoming meetings rounds out the report.

Simulation codes of chemical separation process of spent fuel reprocessing. Tool for process development and safety research

International Nuclear Information System (INIS)

Asakura, Toshihide; Sato, Makoto; Matsumura, Masakazu; Morita, Yasuji

2005-01-01

This paper reviews the succeeding development and utilization of Extraction System Simulation Code for Advanced Reprocessing (ESSCAR). From the viewpoint of development, more tests with spent fuel and calculations should be performed with better understanding of the physico-chemical phenomena in a separation process. From the viewpoint of process safety research on fuel cycle facilities, it is important to know the process behavior of a key substance; being highly reactive but existing only trace amount. (author)

Proceedings of GLOBAL 2007 conference on advanced nuclear fuel cycles and systems

International Nuclear Information System (INIS)

2007-01-01

In keeping with the 12-year history of this conference, GLOBAL 2007 focuses on future nuclear energy systems and fuel cycles. With the increasing public acceptance and political endorsement of nuclear energy, it is a pivotal time for nuclear energy research. Significant advances have been made in development of advanced nuclear fuels and materials, reactor designs, partitioning, transmutation and reprocessing technologies, and waste management strategies. In concert with the technological advances, it is more important than ever to develop sensible nuclear proliferation policies, to promote sustainability, and to continue to increase international collaboration. To further these aims, GLOBAL 2007 highlights recent developments in the following areas: advanced integrated fuel cycle concepts, spent nuclear fuel reprocessing, advanced reprocessing technology, advanced fuels and materials, advanced waste management technology, novel concepts for waste disposal and repository development, advanced reactors, partitioning and transmutation, developments in nuclear non-proliferation technology, policy, and implementation, sustainability and expanded global utilization of nuclear energy, and international collaboration on nuclear energy

Ignition timing advance in the bi-fuel engine

Directory of Open Access Journals (Sweden)

Marek FLEKIEWICZ

2009-01-01

Full Text Available The influence of ignition timing on CNG combustion process has been presented in this paper. A 1.6 liter SI engine has been tested in the special program. For selected engine operating conditions, following data were acquired: in cylinder pressure, crank angle, fuel mass consumption and exhaust gases temperatures. For the timing advance correction varying between 0 to 15 deg crank angle, the internal temperature of combustion chamber, as well as the charge combustion ratio and ratio of heat release has been estimated. With the help of the mathematical model, emissions of NO, CO and CO2 were additionally estimated. Obtained results made it possible to compare the influence of ignition timing advance on natural gas combustion in the SI engine. The engine torque and in-cylinder pressure were used for determination of the optimum engine timing advance.

Application of advanced validation concepts to oxide fuel performance codes: LIFE-4 fast-reactor and FRAPCON thermal-reactor fuel performance codes

Energy Technology Data Exchange (ETDEWEB)

Unal, C., E-mail: cu@lanl.gov [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Williams, B.J. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Yacout, A. [Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Higdon, D.M. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

2013-10-15

Highlights: ► The application of advanced validation techniques (sensitivity, calibration and prediction) to nuclear performance codes FRAPCON and LIFE-4 is the focus of the paper. ► A sensitivity ranking methodology narrows down the number of selected modeling parameters from 61 to 24 for FRAPCON and from 69 to 35 for LIFE-4. ► Fuel creep, fuel thermal conductivity, fission gas transport/release, crack/boundary, and fuel gap conductivity models of LIFE-4 are identified for improvements. ► FRAPCON sensitivity results indicated the importance of the fuel thermal conduction and the fission gas release models. -- Abstract: Evolving nuclear energy programs expect to use enhanced modeling and simulation (M and S) capabilities, using multiscale, multiphysics modeling approaches, to reduce both cost and time from the design through the licensing phases. Interest in the development of the multiscale, multiphysics approach has increased in the last decade because of the need for predictive tools for complex interacting processes as a means of eliminating the limited use of empirically based model development. Complex interacting processes cannot be predicted by analyzing each individual component in isolation. In most cases, the mathematical models of complex processes and their boundary conditions are nonlinear. As a result, the solutions of these mathematical models often require high-performance computing capabilities and resources. The use of multiscale, multiphysics (MS/MP) models in conjunction with high-performance computational software and hardware introduces challenges in validating these predictive tools—traditional methodologies will have to be modified to address these challenges. The advanced MS/MP codes for nuclear fuels and reactors are being developed within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program of the US Department of Energy (DOE) – Nuclear Energy (NE). This paper does not directly address challenges in calibration

Fuel-cycle greenhouse gas emissions impacts of alternative transportation fuels and advanced vehicle technologies

International Nuclear Information System (INIS)

Wang, M. Q.

1998-01-01

At an international conference on global warming, held in Kyoto, Japan, in December 1997, the United States committed to reduce its greenhouse gas (GHG) emissions by 7% over its 1990 level by the year 2012. To help achieve that goal, transportation GHG emissions need to be reduced. Using Argonne's fuel-cycle model, I estimated GHG emissions reduction potentials of various near- and long-term transportation technologies. The estimated per-mile GHG emissions results show that alternative transportation fuels and advanced vehicle technologies can help significantly reduce transportation GHG emissions. Of the near-term technologies evaluated in this study, electric vehicles; hybrid electric vehicles; compression-ignition, direct-injection vehicles; and E85 flexible fuel vehicles can reduce fuel-cycle GHG emissions by more than 25%, on the fuel-cycle basis. Electric vehicles powered by electricity generated primarily from nuclear and renewable sources can reduce GHG emissions by 80%. Other alternative fuels, such as compressed natural gas and liquefied petroleum gas, offer limited, but positive, GHG emission reduction benefits. Among the long-term technologies evaluated in this study, conventional spark ignition and compression ignition engines powered by alternative fuels and gasoline- and diesel-powered advanced vehicles can reduce GHG emissions by 10% to 30%. Ethanol dedicated vehicles, electric vehicles, hybrid electric vehicles, and fuel-cell vehicles can reduce GHG emissions by over 40%. Spark ignition engines and fuel-cell vehicles powered by cellulosic ethanol and solar hydrogen (for fuel-cell vehicles only) can reduce GHG emissions by over 80%. In conclusion, both near- and long-term alternative fuels and advanced transportation technologies can play a role in reducing the United States GHG emissions

Fuel-cycle greenhouse gas emissions impacts of alternative transportation fuels and advanced vehicle technologies.

Energy Technology Data Exchange (ETDEWEB)

Wang, M. Q.

1998-12-16

At an international conference on global warming, held in Kyoto, Japan, in December 1997, the United States committed to reduce its greenhouse gas (GHG) emissions by 7% over its 1990 level by the year 2012. To help achieve that goal, transportation GHG emissions need to be reduced. Using Argonne's fuel-cycle model, I estimated GHG emissions reduction potentials of various near- and long-term transportation technologies. The estimated per-mile GHG emissions results show that alternative transportation fuels and advanced vehicle technologies can help significantly reduce transportation GHG emissions. Of the near-term technologies evaluated in this study, electric vehicles; hybrid electric vehicles; compression-ignition, direct-injection vehicles; and E85 flexible fuel vehicles can reduce fuel-cycle GHG emissions by more than 25%, on the fuel-cycle basis. Electric vehicles powered by electricity generated primarily from nuclear and renewable sources can reduce GHG emissions by 80%. Other alternative fuels, such as compressed natural gas and liquefied petroleum gas, offer limited, but positive, GHG emission reduction benefits. Among the long-term technologies evaluated in this study, conventional spark ignition and compression ignition engines powered by alternative fuels and gasoline- and diesel-powered advanced vehicles can reduce GHG emissions by 10% to 30%. Ethanol dedicated vehicles, electric vehicles, hybrid electric vehicles, and fuel-cell vehicles can reduce GHG emissions by over 40%. Spark ignition engines and fuel-cell vehicles powered by cellulosic ethanol and solar hydrogen (for fuel-cell vehicles only) can reduce GHG emissions by over 80%. In conclusion, both near- and long-term alternative fuels and advanced transportation technologies can play a role in reducing the United States GHG emissions.

Advanced fuel cycles options for LWRs and IMF benchmark definition

International Nuclear Information System (INIS)

Breza, J.; Darilek, P.; Necas, V.

2008-01-01

In the paper, different advanced nuclear fuel cycles including thorium-based fuel and inert-matrix fuel are examined under light water reactor conditions, especially VVER-440, and compared. Two investigated thorium based fuels include one solely plutonium-thorium based fuel and the second one plutonium-thorium based fuel with initial uranium content. Both of them are used to carry and burn or transmute plutonium created in the classical UOX cycle. The inert-matrix fuel consist of plutonium and minor actinides separated from spent UOX fuel fixed in Yttria-stabilised zirconia matrix. The article shows analysed fuel cycles and their short description. The conclusion is concentrated on the rate of Pu transmutation and Pu with minor actinides cumulating in the spent advanced thorium fuel and its comparison to UOX open fuel cycle. Definition of IMF benchmark based on presented scenario is given. (authors)

Physics challenges for advanced fuel cycle assessment

Energy Technology Data Exchange (ETDEWEB)

Giuseppe Palmiotti; Massimo Salvatores; Gerardo Aliberti

2014-06-01

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

Physics challenges for advanced fuel cycle assessment

Energy Technology Data Exchange (ETDEWEB)

Salvatores, Massimo; Aliberti, Gerardo; Palmiotti, Giuseppe

2014-06-17

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

Low Loss Advanced Metallic Fuel Casting Evaluation

International Nuclear Information System (INIS)

Kim, Kihwan; Ko, Youngmo; Kim, Jonghwan; Song, Hoon; Lee Chanbock

2014-01-01

The fabrication process for SFR fuel is composed of fuel slug casting, loading and fabrication of the fuel rods, and the fabrication of the final fuel assemblies. Fuel slug casting is the dominant source of fuel losses and recycles streams in the fabrication process. Recycle streams include fuel slug reworks, returned scraps, and fuel casting heels, which are a special concern in the counter gravity injection casting process because of the large masses involved. Large recycle and waste streams result in lowering the productivity and the economic efficiency of fuel production. To increase efficiency the fuel losses in the furnace chamber, crucible, and the mold, after casting a considerable amount of fuel alloy in the casting furnace, will be quantitatively evaluated. After evaluation the losses will be identified and minimized. It is expected that this study will contribute to the minimization of fuel losses and the wastes streams in the fabrication process of the fuel slugs. Also through this study the technical readiness level of the metallic fuel fabrication process will be further enhanced. In this study, U-Zr alloy system fuel slugs were fabricated by a gravity casting method. Metallic fuel slugs were successfully fabricated with 19 slugs/batch with diameter of 5mm and length of 300mm. Fuel losses was quantitatively evaluated in casting process for the fuel slugs. Fuel losses of the fuel slugs were so low, 0.1∼1.0%. Injection casting experiments have been performed to reduce the fuel loss and improve the casting method. U-Zr fuel slug having φ5.4-L250mm was soundly fabricated with 0.1% in fuel loss. The fuel losses could be minimized to 0.1%, which showed that casting technology of fuel slugs can be a feasible approach to reach the goal of the fuel losses of 0.1% or less in commercial scale

ABB advanced BWR and PWR fuel

International Nuclear Information System (INIS)

Junkrans, S.; Helmersson, S.; Andersson, S.

1999-01-01

Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

A catalogue of advanced fuel cycles in CANDU-PHW reactors

International Nuclear Information System (INIS)

Veeder, J.; Didsbury, R.

1985-06-01

A catalogue raisonne is presented of various advanced fuel cycle options which have the potential of substantially improving the uranium utilization for CANDU-PHW reactors. Three categories of cycles are: once-through cycles without recovery of fissile materials, cycles that depend on the recovery and recycle of fissile materials in thorium or uranium, cycles that depend primarily on the production of fissile material in a fertile blanket by means of an intense neutron source other than fission, such as an accelerator breeder. Detailed tables are given of the isotopic compositions of the feed and discharge fuels, the logistics of materials and processes required to sustain each of the cycles, and tables of fuel cycle costs based on a method of continuous discounting of cash flow

The status of nuclear fuel cycle system analysis for the development of advanced nuclear fuel cycles

Energy Technology Data Exchange (ETDEWEB)

Ko, Won Il; Kim, Seong Ki; Lee, Hyo Jik; Chang, Hong Rae; Kwon, Eun Ha; Lee, Yoon Hee; Gao, Fanxing [KAERI, Daejeon (Korea, Republic of)

2011-11-15

The system analysis has been used with different system and objectives in various fields. In the nuclear field, the system can be applied from uranium mining to spent fuel reprocessing or disposal which is called the nuclear fuel cycle. The analysis of nuclear fuel cycle can be guideline for development of advanced fuel cycle through integrating and evaluating the technologies. For this purpose, objective approach is essential and modeling and simulation can be useful. In this report, several methods which can be applicable for development of advanced nuclear fuel cycle, such as TRL, simulation and trade analysis were explained with case study

Advanced chemical hydride-based hydrogen generation/storage system for fuel cell vehicles

Energy Technology Data Exchange (ETDEWEB)

Breault, R.W.; Rolfe, J. [Thermo Power Corp., Waltham, MA (United States)

1998-08-01

Because of the inherent advantages of high efficiency, environmental acceptability, and high modularity, fuel cells are potentially attractive power supplies. Worldwide concerns over clean environments have revitalized research efforts on developing fuel cell vehicles (FCV). As a result of intensive research efforts, most of the subsystem technology for FCV`s are currently well established. These include: high power density PEM fuel cells, control systems, thermal management technology, and secondary power sources for hybrid operation. For mobile applications, however, supply of hydrogen or fuel for fuel cell operation poses a significant logistic problem. To supply high purity hydrogen for FCV operation, Thermo Power`s Advanced Technology Group is developing an advanced hydrogen storage technology. In this approach, a metal hydride/organic slurry is used as the hydrogen carrier and storage media. At the point of use, high purity hydrogen will be produced by reacting the metal hydride/organic slurry with water. In addition, Thermo Power has conceived the paths for recovery and regeneration of the spent hydride (practically metal hydroxide). The fluid-like nature of the spent hydride/organic slurry will provide a unique opportunity for pumping, transporting, and storing these materials. The final product of the program will be a user-friendly and relatively high energy storage density hydrogen supply system for fuel cell operation. In addition, the spent hydride can relatively easily be collected at the pumping station and regenerated utilizing renewable sources, such as biomass, natural, or coal, at the central processing plants. Therefore, the entire process will be economically favorable and environmentally friendly.

Development of advanced fabrication technology for high-temperature gas-cooled reactor fuel. Reduction of coating failure fraction

International Nuclear Information System (INIS)

Minato, Kazuo; Kikuchi, Hironobu; Fukuda, Kousaku; Tobita, Tsutomu; Yoshimuta, Sigeharu; Suzuki, Nobuyuki; Tomimoto, Hiroshi; Nishimura, Kazuhisa; Oda, Takafumi

1998-11-01

The advanced fabrication technology for high-temperature gas-cooled reactor fuel has been developed to reduce the coating failure fraction of the fuel particles, which leads to an improvement of the reactor safety. The present report reviews the results of the relevant work. The mechanisms of the coating failure of the fuel particles during coating and compaction processes of the fuel fabrication were studied to determine a way to reduce the coating failure fraction of the fuel. The coating process was improved by optimizing the mode of the particle fluidization and by developing the process without unloading and loading of the particles at intermediate coating process. The compaction process was improved by optimizing the combination of the pressing temperature and the pressing speed of the overcoated particles. Through these modifications of the fabrication process, the quality of the fuel was improved outstandingly. (author)

Advanced Fuel Cell System Thermal Management for NASA Exploration Missions

Science.gov (United States)

Burke, Kenneth A.

2009-01-01

The NASA Glenn Research Center is developing advanced passive thermal management technology to reduce the mass and improve the reliability of space fuel cell systems for the NASA exploration program. An analysis of a state-of-the-art fuel cell cooling systems was done to benchmark the portion of a fuel cell system s mass that is dedicated to thermal management. Additional analysis was done to determine the key performance targets of the advanced passive thermal management technology that would substantially reduce fuel cell system mass.

Radiation and physical protection challenges at advanced nuclear fuel cycle facilities

International Nuclear Information System (INIS)

Pickett, Susan E.

2008-01-01

Full text: The purpose of this study is to examine challenges and opportunities for radiation protection in advanced nuclear reactors and fuel facilities proposed under the Generation IV (GEN IV) initiative which is examining and pursuing the exploration and development of advanced nuclear science and technology; and the Global Nuclear Energy Partnership (GNEP), which seeks to develop worldwide consensus on enabling expanded use of economical, carbon-free nuclear energy to meet growing energy demand. The International Energy Agency projects nuclear power to increase at a rate of 1.3 to 1.5 percent a year over the next 20 years, depending on economic growth. Much of this growth will be in Asia, which, as a whole, currently has plans for 40 new nuclear power plants. Given this increase in demand for new nuclear power facilities, ranging from light water reactors to advanced fuel processing and fabrication facilities, it is necessary for radiation protection and physical protection technologies to keep pace to ensure both worker and public health. This paper is based on a review of current initiatives and the proposed reactors and facilities, primarily the nuclear fuel cycle facilities proposed under the GEN IV and GNEP initiatives. Drawing on the Technology Road map developed under GEN IV, this work examines the potential radiation detection and protection challenges and issues at advanced reactors, including thermal neutron spectrum systems, fast neutron spectrum systems and nuclear fuel recycle facilities. The thermal neutron systems look to improve the efficiency of production of hydrogen or electricity, while the fast neutron systems aim to enable more effective management of actinides through recycling of most components in the discharged fuel. While there are components of these advanced systems that can draw on the current and well-developed radiation protection practices, there will inevitably be opportunities to improve the overall quality of radiation

ABB Turbo advanced fuel for application in System 80 family of plants

International Nuclear Information System (INIS)

Karoutas, Z.E.; Dixon, D.J.; Shapiro, N.L.

1998-01-01

ABB Combustion Engineering Nuclear Operations (ABB CE) has developed an Advanced Fuel Design, tailored to the Combustion Engineering, Inc. (CE) Nuclear Steam Supply System (NSSS) environment. This Advanced Fuel Design called Turbo features a full complement of innovative components, including GUARDIAN debris-resistant spacer grids, Turbo Zircaloy mixing grids to increase thermal margin and grid-to-rod fretting resistance, value-added fuel pellets to increase fuel loading, advanced cladding to increase achievable burnup, and axial blankets and Erbium integral burnable absorbers for improving fuel cycle economics. This paper summarizes the Turbo Fuel Design and its application to a System 80 family type plant. Benefits in fuel reliability, thermal margin, improved fuel cycle economics and burn up capability are compared relative to the current ABB CE standard fuel design. The fuel management design and the associated thermal margin are also evaluated. (author)

Radioactive waste management of experimental DUPIC fuel fabrication process

International Nuclear Information System (INIS)

Lee, H. H.; Park, J. J.; Shin, J. M.; Yang, M. S.; Hong, K. P.

2001-01-01

The concept of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) is a dry processing technology to manufacture CANDU compatible DUPIC fuel from spent PWR fuel material. Real spent PWR fuel was used in IMEF M6 hot cell to carry out DUPIC experiment. Afterwards, about 200 kg-U of spent PWR fuel is supposed to be used till 2006. This study has been conducted in some hot cells of PIEF and M6 cell of IMEF. There are various forms of nuclear material such as rod cut, powder, green pellet, sintered pellet, fabrication debris, fuel rod, fuel bundle, sample, and process waste produced from various manufacturing experiment of DUPIC fuel. After completing test, the above nuclear wastes and test equipment etc. will be classified as radioactive waste, transferred to storage facility and managed rigorously according to domestic and international laws until the final management policy is determined. It is desirable to review management options in advance for radioactive waste generated from manufacturing experiment of DUPIC nuclear fuel as well as residual nuclear material and dismantled equipment. This paper includes basic plan for DUPIC radwaste, arising source and estimated amount of radioactive waste, waste classification and packing, transport cask, transport procedures

Safety issues on advanced fuel

Energy Technology Data Exchange (ETDEWEB)

Gross, H.; Krebs, W.D. [Siemens AG, Bereich Energieerzeugug (KWU), Erlangen (Germany). Geschaeftsgebiet Nukleare Energieerzeugung

1998-05-01

In the recent years a general discussion has started whether unsolved safety issues are related to advanced fuel. Advanced fuel is in this context a summary of features like high burnup, improved clad materials, low leakage loading pattern with high peaking factors etc. The design basis accidents RIA and Loca are of special interest for this discussion. From the Siemens point of view RIA is not a safety issue. There are sufficient margins between the enthalpy rise calculated by modern 3D methods and the fuel failures which occurred in RIA simulation tests when the effect of pulse width is taken into account. The evaluation of possible uncertainties for the established Loca criteria (17% equivalent corrosion, 1200 C clad temperature) for high burnup makes sense. But fuel with high burnup has significantly lower peaking factors than fuel with lower burnup. This gives sufficient margin counterbalancing possible uncertainties. In contrast to the above incomplete control rod insertion at higher burnup is potentially a real safety issue. Although Siemens fuel was not affected by the reported incidents they addressed the problem and checked that they have sufficient design margin for their fuel. (orig.) [Deutsch] In den letzten Jahren hat eine allgemeine Diskussion begonnen, ob mit fortgeschrittenen Brennelementen (BE) ungeklaerte Sicherheitsprobleme verbunden sind. Dabei ist `Fortgeschrittene Brennelemente` ein Sammelbegriff fuer hohe Abbraende, verbesserte Huellrohrmaterialien, Low-leakage-Einsatzplanungen mit hohen Heissstellenfaktoren usw. Die Auslegungsstoerfaelle RIA und Loca sind in dieser Diskussion von besonderer Bedeutung. Aus der Sicht von Siemens ist der RIA kein Sicherheitsproblem. Zwischen den mit modernen 3D-Methoden berechneten Enthalpieerhoehungen und den in RIA-Experimenten aufgetretenen Brennstabdefekten bestehen ausreichende Abstaende, wenn der Einfluss der Pulsbreite beruecksichtigt wird. Die Untersuchung eventueller Unsicherheiten bei hohen

Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor

International Nuclear Information System (INIS)

Ott, Larry J.; Ellis, Ronald James; McDuffee, Joel Lee; Spellman, Donald J.; Bevard, Bruce Balkcom

2009-01-01

The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions (i.e., for both fast-spectrum conditions and light-water-reactor conditions). This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.

Advances in medium and high temperature solid oxide fuel cell technology

CERN Document Server

Salvatore, Aricò

2017-01-01

In this book well-known experts highlight cutting-edge research priorities and discuss the state of the art in the field of solid oxide fuel cells giving an update on specific subjects such as protonic conductors, interconnects, electrocatalytic and catalytic processes and modelling approaches. Fundamentals and advances in this field are illustrated to help young researchers address issues in the characterization of materials and in the analysis of processes, not often tackled in scholarly books.

Advances in High Temperature Gas Cooled Reactor Fuel Technology

International Nuclear Information System (INIS)

2012-12-01

This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

Advances in High Temperature Gas Cooled Reactor Fuel Technology

International Nuclear Information System (INIS)

2012-06-01

This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

PEM Fuel Cells with Bio-Ethanol Processor Systems A Multidisciplinary Study of Modelling, Simulation, Fault Diagnosis and Advanced Control

CERN Document Server

Feroldi, Diego; Outbib, Rachid

2012-01-01

An apparently appropriate control scheme for PEM fuel cells may actually lead to an inoperable plant when it is connected to other unit operations in a process with recycle streams and energy integration. PEM Fuel Cells with Bio-Ethanol Processor Systems presents a control system design that provides basic regulation of the hydrogen production process with PEM fuel cells. It then goes on to construct a fault diagnosis system to improve plant safety above this control structure. PEM Fuel Cells with Bio-Ethanol Processor Systems is divided into two parts: the first covers fuel cells and the second discusses plants for hydrogen production from bio-ethanol to feed PEM fuel cells. Both parts give detailed analyses of modeling, simulation, advanced control, and fault diagnosis. They give an extensive, in-depth discussion of the problems that can occur in fuel cell systems and propose a way to control these systems through advanced control algorithms. A significant part of the book is also given over to computer-aid...

National Jet Fuels Combustion Program – Area #3 : Advanced Combustion Tests

Science.gov (United States)

2017-12-31

The goal of this study is to develop, conduct, and analyze advanced laser and optical measurements in the experimental combustors developed under ASCENT National Fuel Combustion Program to measure sensitivity to fuel properties. We conducted advanced...

Development of Advanced High Uranium Density Fuels for Light Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Blanchard, James [Univ. of Wisconsin, Madison, WI (United States); Butt, Darryl [Boise State Univ., ID (United States); Meyer, Mitchell [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Corporation, Pittsburgh, PA (United States)

2016-02-15

This work conducts basic materials research (fabrication, radiation resistance, thermal conductivity, and corrosion response) on U3Si2 and UN, two high uranium density fuel forms that have a high potential for success as advanced light water reactor (LWR) fuels. The outcome of this proposed work will serve as the basis for the development of advance LWR fuels, and utilization of such fuel forms can lead to the optimization of the fuel performance related plant operating limits such as power density, power ramp rate and cycle length.

Results of modeling advanced BWR fuel designs using CASMO-4

International Nuclear Information System (INIS)

Knott, D.; Edenius, M.

1996-01-01

Advanced BWR fuel designs from General Electric, Siemens and ABB-Atom have been analyzed using CASMO-4 and compared against fission rate distributions and control rod worths from MCNP. Included in the analysis were fuel storage rack configurations and proposed mixed oxide (MOX) designs. Results are also presented from several cycles of SIMULATE-3 core follow analysis, using nodal data generated by CASMO-4, for cycles in transition from 8x8 designs to advanced fuel designs. (author)

Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview

Energy Technology Data Exchange (ETDEWEB)

Unal, Cetin [Los Alamos National Laboratory; Pasamehmetoglu, Kemal [IDAHO NATIONAL LAB; Carmack, Jon [IDAHO NATIONAL LAB

2010-01-01

Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R & D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview

International Nuclear Information System (INIS)

Unal, Cetin; Pasamehmetoglu, Kemal; Carmack, Jon

2010-01-01

Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R and D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

An Investigation on Irradiation-induced Grid Width Growth in Advanced Fuels

International Nuclear Information System (INIS)

Jang, Young Ki; Jeon, Kyeong Lak; Kim, Yong Hwan; Kim, Jae Ik; Hwang, Sun Tack; Kim, Man Su; Lee, Tae Hyoung; Yoo, Myeong Jong; Yoon, Yong Bae; Kim, Tae Wan

2011-01-01

The spacer grids for fuel assembly are fabricated from preformed Zircaloy or Inconel strips interlocked in an egg crate fashion and welded or brazed together. The spacer grid is the important component to maintain the fuel rod array by providing positive lateral restraint to the fuel rods but only frictional restraint to axial fuel rod motion. To improve economy and safety aspects, advanced nuclear fuels of PLUS7, 16ACE7 and 17ACE7 were developed. The former is for Optimized Power Reactor of 1000 MWe (OPR1000) and Advanced Power Reactor of 1400 MWe (APR1400) and the latter two are for 16x16 and 17x17 Westinghouse type reactors, respectively. The material for top and bottom spacer grids on these advanced fuels are Inconel and the mid grids are Zirlo patented by Westinghouse. For neutron economy, the fuel assemblies are arranged very closely and the gaps between assemblies are kept to around 1 mm based on the worst case. The Zirconium-based alloys grow during irradiation in reactor. The large growth may cause some difficulties in loading and unloading fuel assemblies during refueling outage in reactor. The severe growth may cause some problems that fuel assemblies may be stuck within the core shroud and a modification of loading pattern is required. In addition, the grid growth with grid spring relaxation may cause different rod vibration behavior and results in the different wear mechanism. The grid width growth on the advanced fuels were predicted by using the growth models before the irradiation in reactor and were examined using lead test assemblies (LTAs) after each cycle in Ulchin unit 3 and Kori units 2 and 3, respectively. To reconfirm irradiation performance results using LTAs, the additional examinations are being performed through the surveillance programs on the commercially supplied fuels in Yonggwang unit 5 and Kori units 2 and 4. It is investigated on this study whether the grid widths on the advanced fuels meet their criteria and the predicted models

Construction and engineering report for advanced nuclear fuel development facility

International Nuclear Information System (INIS)

Cho, S. W.; Park, J. S.; Kwon, S.J.; Lee, K. W.; Kim, I. J.; Yu, C. H.

2003-09-01

The design and construction of the fuel technology development facility was aimed to accommodate general nuclear fuel research and development for the HANARO fuel fabrication and advanced fuel researches. 1. Building size and room function 1) Building total area : approx. 3,618m 2 , basement 1st floor, ground 3th floor 2) Room function : basement floor(machine room, electrical room, radioactive waste tank room), 1st floor(research reactor fuel fabrication facility, pyroprocess lab., metal fuel lab., nondestructive lab., pellet processing lab., access control room, sintering lab., etc), 2nd floor(thermal properties measurement lab., pellet characterization lab., powder analysis lab., microstructure analysis lab., etc), 3rd floor(AHU and ACU Room) 2. Special facility equipment 1) Environmental pollution protection equipment : ACU(2sets), 2) Emergency operating system : diesel generator(1set), 3) Nuclear material handle, storage and transport system : overhead crane(3sets), monorail hoist(1set), jib crane(2sets), tank(1set) 4) Air conditioning unit facility : AHU(3sets), packaged air conditioning unit(5sets), 5) Automatic control system and fire protection system : central control equipment(1set), lon device(1set), fire hose cabinet(3sets), fire pump(3sets) etc

Pyrometallurgical processing of Integral Fast Reactor metal fuels

International Nuclear Information System (INIS)

Battles, J.E.; Miller, W.E.; Gay, E.C.

1991-01-01

The pyrometallurgical process for recycling spent metal fuels from the Integral Fast Reactor is now in an advanced state of development. This process involves electrorefining spent fuel with a cadmium anode, solid and liquid cathodes, and a molten salt electrolyte (LiCl-KCl) at 500 degrees C. The initial process feasibility and flowsheet verification studies have been conducted in a laboratory-scale electrorefiner. Based on these studies, a dual cathode approach has been adopted, where uranium is recovered on a solid cathode mandrel and uranium-plutonium is recovered in a liquid cadmium cathode. Consolidation and purification (salt and cadmium removal) of uranium and uranium-plutonium products from the electrorefiner have been successful. The process is being developed with the aid of an engineering-scale electrorefiner, which has been successfully operated for more than three years. In this electrorefiner, uranium has been electrotransported from the cadmium anode to a solid cathode in 10 kg quantities. Also, anodic dissolution of 10 kg batches of chopped, simulated fuel (U--10% Zr) has been demonstrated. Development of the liquid cadmium cathode for recovering uranium-plutonium is under way

Feasibility study on the development of advanced LWR fuel technology

International Nuclear Information System (INIS)

Jung, Youn Ho; Sohn, D. S.; Jeong, Y. H.; Song, K. W.; Song, K. N.; Chun, T. H.; Bang, J. G.; Bae, K. K.; Kim, D. H. and others.

1997-07-01

Worldwide R and D trends related to core technology of LWR fuels and status of patents have been surveyed for the feasibility study. In addition, various fuel cycle schemes have been studied to establish the target performance parameters. For the development of cladding material, establishment of long-term research plan for alloy development and optimization of melting process and manufacturing technology were conducted. A work which could characterize the effect of sintering additives on the microstructure of UO 2 pellet has been experimentally undertaken, and major sintering variables and their ranges have been found in the sintering process of UO 2 -Gd 2 O 3 burnable absorber pellet. The analysis of state of the art technology related to flow mixing device for spacer grid and debris filtering device for bottom nozzle and the investigation of the physical phenomena related to CHF enhancement and the establishment of the data base for thermal-hydraulic performance tests has been done in this study. In addition, survey on the documents of the up-to-date PWR fuel assemblies developed by foreign vendors have been carried out to understand their R and D trends and establish the direction of R and D for these structural components. And, to set the performance target of the new fuel, to be developed, fuel burnup and economy under the extended fuel cycle length scheme were estimated. A preliminary study on the failure mechanism of CANDU fuel, key technology and advanced coating has been performed. (author). 190 refs., 31 tabs., 129 figs

Feasibility study on the development of advanced LWR fuel technology

Energy Technology Data Exchange (ETDEWEB)

Jung, Youn Ho; Sohn, D. S.; Jeong, Y. H.; Song, K. W.; Song, K. N.; Chun, T. H.; Bang, J. G.; Bae, K. K.; Kim, D. H. and others

1997-07-01

Worldwide R and D trends related to core technology of LWR fuels and status of patents have been surveyed for the feasibility study. In addition, various fuel cycle schemes have been studied to establish the target performance parameters. For the development of cladding material, establishment of long-term research plan for alloy development and optimization of melting process and manufacturing technology were conducted. A work which could characterize the effect of sintering additives on the microstructure of UO{sub 2} pellet has been experimentally undertaken, and major sintering variables and their ranges have been found in the sintering process of UO{sub 2}-Gd{sub 2}O{sub 3} burnable absorber pellet. The analysis of state of the art technology related to flow mixing device for spacer grid and debris filtering device for bottom nozzle and the investigation of the physical phenomena related to CHF enhancement and the establishment of the data base for thermal-hydraulic performance tests has been done in this study. In addition, survey on the documents of the up-to-date PWR fuel assemblies developed by foreign vendors have been carried out to understand their R and D trends and establish the direction of R and D for these structural components. And, to set the performance target of the new fuel, to be developed, fuel burnup and economy under the extended fuel cycle length scheme were estimated. A preliminary study on the failure mechanism of CANDU fuel, key technology and advanced coating has been performed. (author). 190 refs., 31 tabs., 129 figs.

Polarized advanced fuel reactors

International Nuclear Information System (INIS)

Kulsrud, R.M.

1987-07-01

The d- 3 He reaction has the same spin dependence as the d-t reaction. It produces no neutrons, so that if the d-d reactivity could be reduced, it would lead to a neutron-lean reactor. The current understanding of the possible suppression of the d-d reactivity by spin polarization is discussed. The question as to whether a suppression is possible is still unresolved. Other advanced fuel reactions are briefly discussed. 11 refs

Research on advanced aqueous reprocessing of spent nuclear fuel: literature study

Energy Technology Data Exchange (ETDEWEB)

Van Hecke, K.; Goethals, P.

2006-07-15

The goal of the partitioning and transmutation strategy is to reduce the radiotoxicity of spent nuclear fuel to the level of natural uranium in a short period of time (about 1000 years) and thus the required containment period of radioactive material in a repository. Furthermore, it aims to reduce the volume of waste requiring deep geological disposal and hence the associated space requirements and costs. Several aqueous as well as pyrochemical separation processes have been developed for the partitioning of the long-lived radionuclides from the remaining of the spent fuel. This report aims to describe and compare advanced aqueous reprocessing methods.

Research on advanced aqueous reprocessing of spent nuclear fuel: literature study

International Nuclear Information System (INIS)

Van Hecke, K.; Goethals, P.

2006-01-01

The goal of the partitioning and transmutation strategy is to reduce the radiotoxicity of spent nuclear fuel to the level of natural uranium in a short period of time (about 1000 years) and thus the required containment period of radioactive material in a repository. Furthermore, it aims to reduce the volume of waste requiring deep geological disposal and hence the associated space requirements and costs. Several aqueous as well as pyrochemical separation processes have been developed for the partitioning of the long-lived radionuclides from the remaining of the spent fuel. This report aims to describe and compare advanced aqueous reprocessing methods.

Operating experience with Exxon nuclear advanced fuel assembly and fuel cycle designs in PWRs

International Nuclear Information System (INIS)

Skogen, F.B.; Killgore, M.R.; Holm, J.S.; Brown, C.A.

1986-01-01

Exxon Nuclear Company (ENC) has achieved a high standard of performance in its supply of fuel reloads for both BWRs and PWRs, while introducing substantial innovations aimed at realization of improved fuel cycle costs. The ENC experience with advanced design features such as the bi-metallic spacer, the dismountable upper tie plate, natural uranium axial blankets, optimized water-to-fuel designs, annular pellets, gadolinia burnable absorbers, and improved fuel management scenarios, is summarized

Advances in HTGR spent fuel treatment technology

International Nuclear Information System (INIS)

Holder, N.D.; Lessig, W.S.

1984-08-01

GA Technologies, Inc. has been investigating the burning of spent reactor graphite under Department of Energy sponsorship since 1969. Several deep fluidized bed burners have been used at the GA pilot plant to develop graphite burning techniques for both spent fuel recovery and volume reduction for waste disposal. Since 1982 this technology has been extended to include more efficient circulating bed burners. This paper includes updates on high-temperature gas-cooled reactor fuel cycle options and current results of spent fuel treatment testing for fluidized and advanced circulating bed burners

Fuel cells science and engineering. Materials, processes, systems and technology. Vol. 1

Energy Technology Data Exchange (ETDEWEB)

Stolten, Detlef; Emonts, Bernd (eds.) [Forschungszentrum Juelich GmbH (DE). Inst. fuer Energieforschung (IEF), Brennstoffzellen (IEF-3)

2012-07-01

The first volume is divided in four parts and 22 chapters. It is structured as follows: PART I: Technology. Chapter 1: Technical Advancement of Fuel-Cell Research and Development (Dr. Bernd Emonts, Ludger Blum, Thomas Grube, Werner Lehnert, Juergen Mergel, Martin Mueller and Ralf Peters); 2: Single-Chamber Fuel Cells (Teko W. Napporn and Melanie Kuhn); 3: Technology and Applications of Molten Carbonate Fuel Cells (Barbara Bosio, Elisabetta Arato and Paolo Greppi); 4: Alkaline Fuel Cells (Erich Guelzow); 5: Micro Fuel Cells (Ulf Groos and Dietmar Gerteisen); 6: Principles and Technology of Microbial Fuel Cells (Jan B. A. Arends, Joachim Desloover, Sebastia Puig and Willy Verstraete); 7: Micro-Reactors for Fuel Processing (Gunther Kolb); 8: Regenerative Fuel Cells (Martin Mueller). PART II: Materials and Production Processes. Chapter 9: Advances in Solid Oxide Fuel Cell Development between 1995 and 2010 at Forschungszentrum Juelich GmbH, Germany (Vincent Haanappel); 10: Solid Oxide Fuel Cell Electrode Fabrication by Infiltration (Evren Gunen); 11: Sealing Technology for Solid Oxide Fuel Cells (K. Scott Weil); 12: Phosphoric Acid, an Electrolyte for Fuel Cells - Temperature and Composition Dependence of Vapor Pressure and Proton Conductivity (Carsten Korte); 13: Materials and Coatings for Metallic Bipolar Plates in Polymer Electrolyte Membrane Fuel Cells (Heli Wang and John A. Turner); 14: Nanostructured Materials for Fuel Cells (John F. Elter); 15: Catalysis in Low-Temperature Fuel Cells - An Overview (Sabine Schimpf and Michael Bron). PART III: Analytics and Diagnostics. Chapter 16: Impedance Spectroscopy for High-Temperature Fuel Cells (Ellen Ivers-Tiffee, Andre Leonide, Helge Schichlein, Volker Sonn and Andre Weber); 17: Post-Test Characterization of Solid Oxide Fuel-Cell Stacks (Norbert H. Menzler and Peter Batfalsky); 18: In Situ Imaging at Large-Scale Facilities (Christian Toetzke, Ingo Manke and Werner Lehnert); 19: Analytics of Physical Properties of Low

Shield wall evaluation of hot cell facility for advanced spent fuel conditioning process

International Nuclear Information System (INIS)

Cho, I. J.; Kuk, D. H.; Ko, J. H.; Jung, W. M.; Yoo, G. S.; Lee, E. P.; Park, S. W.

2002-01-01

The future hot cell is located in the Irradiated Material Experiment Facility (IMEF) at the Korea Atomic Energy Research Institute (KAERI). It is β-γ type hot cell that was constructed on the base floor in IMEF building for irradiated material testing. And this hot cell will be used for carrying out the Advanced spent fuel Conditioning Process (ACP). The radiation shielding capability of hot cell should be sufficient to meet the radiation dose requirements in the related regulations. Because the radioactive sources of ACP are expected to be higher than radioactive sources of IMEF design criteria, the future hot cell in current status is unsatisfactory to hot test of ACP. So the shielding analysis of the future hot cell is performed to evaluate shielding ability of concrete shield wall. The shielding analysis included (a) identification of ACP source term; (b) photon source spectrum; (c) shielding analysis by QADS and MCNP-4C; and (d) enhancement of concrete shield wall. In this research, dose rates are obtained according to ACP source, geometry and hot cell shield wall thickness. And the evaluation and reinforcement thickness of the shield wall about future hot cell are concluded

Use of freeze-casting in advanced burner reactor fuel design

Energy Technology Data Exchange (ETDEWEB)

Lang, A. L.; Yablinsky, C. A.; Allen, T. R. [Dept. of Engineering Physics, Univ. of Wisconsin Madison, 1500 Engineering Drive, Madison, WI 53711 (United States); Burger, J.; Hunger, P. M.; Wegst, U. G. K. [Thayer School of Engineering, Dartmouth College, 8000 Cummings Hall, Hanover, NH 03755 (United States)

2012-07-01

This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by that fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO{sub 2}) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models

Advanced technique for computing fuel combustion properties in pulverized-fuel fired boilers

Energy Technology Data Exchange (ETDEWEB)

Kotler, V.R. (Vsesoyuznyi Teplotekhnicheskii Institut (Russian Federation))

1992-03-01

Reviews foreign technical reports on advanced techniques for computing fuel combustion properties in pulverized-fuel fired boilers and analyzes a technique developed by Combustion Engineering, Inc. (USA). Characteristics of 25 fuel types, including 19 grades of coal, are listed along with a diagram of an installation with a drop tube furnace. Characteristics include burn-out intensity curves obtained using thermogravimetric analysis for high-volatile bituminous, semi-bituminous and coking coal. The patented LFP-SKM mathematical model is used to model combustion of a particular fuel under given conditions. The model allows for fuel particle size, air surplus, load, flame height, and portion of air supplied as tertiary blast. Good agreement between computational and experimental data was observed. The method is employed in designing new boilers as well as converting operating boilers to alternative types of fuel. 3 refs.

Experience with advanced driver fuels in EBR-II

International Nuclear Information System (INIS)

Lahm, C.E.; Koenig, J.F.; Pahl, R.G.; Porter, D.L.; Crawford, D.C.

1992-01-01

This paper discusses several metallic fuel element designs which have been tested and used as driver fuel in Experimental Breeder Reactor II (EBR-II). The most recent advanced designs have all performed acceptably in EBR-H and can provide reliable performance to high burnups. Fuel elements tested have included use of U-l0Zr metallic fuel with either D9, 316 or HT9 stainless steel cladding; the D9 and 316-clad designs have been used as standard driver fuel. Experimental data indicate that fuel performance characteristics are very similar for the various designs tested. Cladding materials can be selected that optimize performance based on reactor design and operational goals

Recovery Act: Advanced Direct Methanol Fuel Cell for Mobile Computing

Energy Technology Data Exchange (ETDEWEB)

Fletcher, James H. [University of North Florida; Cox, Philip [University of North Florida; Harrington, William J [University of North Florida; Campbell, Joseph L [University of North Florida

2013-09-03

containment. PROJECT OVERVIEW The University of North Florida (UNF), with project partner the University of Florida, recently completed the Department of Energy (DOE) project entitled “Advanced Direct Methanol Fuel Cell for Mobile Computing”. The primary objective of the project was to advance portable fuel cell system technology towards the commercial targets as laid out in the DOE R&D roadmap by developing a 20-watt, direct methanol fuel cell (DMFC), portable power supply based on the UNF innovative “passive water recovery” MEA. Extensive component, sub-system, and system development and testing was undertaken to meet the rigorous demands of the consumer electronic application. Numerous brassboard (nonpackaged) systems were developed to optimize the integration process and facilitating control algorithm development. The culmination of the development effort was a fully-integrated, DMFC, power supply (referred to as DP4). The project goals were 40 W/kg for specific power, 55 W/l for power density, and 575 Whr/l for energy density. It should be noted that the specific power and power density were for the power section only, and did not include the hybrid battery. The energy density is based on three, 200 ml, fuel cartridges, and also did not include the hybrid battery. The results show that the DP4 system configured without the methanol concentration sensor exceeded all performance goals, achieving 41.5 W/kg for specific power, 55.3 W/l for power density, and 623 Whr/l for energy density. During the project, the DOE revised its technical targets, and the definition of many of these targets, for the portable power application. With this revision, specific power, power density, specific energy (Whr/kg), and energy density are based on the total system, including fuel tank, fuel, and hybridization battery. Fuel capacity is not defined, but the same value is required for all calculations. Test data showed that the DP4 exceeded all 2011 Technical Status values; for example

Testing Systems and Results for Advanced Nuclear Fuel Materials

International Nuclear Information System (INIS)

Rooyen, I.J. van; Griffith, G.W.; Garnier, J.E.

2012-01-01

Light Water Reactor Sustainability (LWRS) Program Advanced LWR Nuclear Fuel Development (ALFD) Pathway. Development and testing of high performance fuel cladding identified as high priority to support: enhancement of fuel performance, reliability, and reactor safety. One of the technologies being examined is an advanced fuel cladding made from ceramic matrix composites (CMC) utilizing silicon carbide (SiC) as a structural material supplementing a commercial Zircaloy-4 (Zr-4) tube. A series of out-of-pile tests to fully characterize the SiC CMC hybrid design to produce baseline data. The planned tests are intended to either produce quantitative data or to demonstrate the properties required to achieve two initial performance conditions relative to standard zircaloybased cladding: decreased hydrogen uptake (corrosion) and decreased fretting of the cladding tube under normal operating and postulated accident conditions. These two failure mechanisms account for approximately 70% of all in-pile failures of LWR commercial fuel assemblies

Implications of alpha-decay for long term storage of advanced heavy water reactor fuels

International Nuclear Information System (INIS)

Pencer, J.; McDonald, M.H.; Roubtsov, D.; Edwards, G.W.R.

2017-01-01

Highlights: •Alpha decays versus storage time are calculated for examples of advanced heavy water reactor fuels. •Estimates are made for fuel swelling and helium bubble formation as a function of time. •These predictions are compared to predictions for natural uranium fuel. •Higher rates of damage are predicted for advanced heavy water reactor fuels than natural uranium. -- Abstract: The decay of actinides such as 238 Pu, results in recoil damage and helium production in spent nuclear fuels. The extent of the damage depends on storage time and spent fuel composition and has implications for the integrity of the fuels. Some advanced nuclear fuels intended for use in pressurized heavy water pressure tube reactors have high initial plutonium content and are anticipated to exhibit swelling and embrittlement, and to accumulate helium bubbles over storage times as short as hundreds of years. Calculations are performed to provide estimates of helium production and fuel swelling associated with alpha decay as a function of storage time. Significant differences are observed between predicted aging characteristics of natural uranium and the advanced fuels, including increased helium concentrations and accelerated fuel swelling in the latter. Implications of these observations for long term storage of advanced fuels are discussed.

ADVANCING THE FUNDAMENTAL UNDERSTANDING AND SCALE-UP OF TRISO FUEL COATERS VIA ADVANCED MEASUREMENT AND COMPUTATIONAL TECHNIQUES

Energy Technology Data Exchange (ETDEWEB)

Biswas, Pratim; Al-Dahhan, Muthanna

2012-11-01

Tri-isotropic (TRISO) fuel particle coating is critical for the future use of nuclear energy produced byadvanced gas reactors (AGRs). The fuel kernels are coated using chemical vapor deposition in a spouted fluidized bed. The challenges encountered in operating TRISO fuel coaters are due to the fact that in modern AGRs, such as High Temperature Gas Reactors (HTGRs), the acceptable level of defective/failed coated particles is essentially zero. This specification requires processes that produce coated spherical particles with even coatings having extremely low defect fractions. Unfortunately, the scale-up and design of the current processes and coaters have been based on empirical approaches and are operated as black boxes. Hence, a voluminous amount of experimental development and trial and error work has been conducted. It has been clearly demonstrated that the quality of the coating applied to the fuel kernels is impacted by the hydrodynamics, solids flow field, and flow regime characteristics of the spouted bed coaters, which themselves are influenced by design parameters and operating variables. Further complicating the outlook for future fuel-coating technology and nuclear energy production is the fact that a variety of new concepts will involve fuel kernels of different sizes and with compositions of different densities. Therefore, without a fundamental understanding the underlying phenomena of the spouted bed TRISO coater, a significant amount of effort is required for production of each type of particle with a significant risk of not meeting the specifications. This difficulty will significantly and negatively impact the applications of AGRs for power generation and cause further challenges to them as an alternative source of commercial energy production. Accordingly, the proposed work seeks to overcome such hurdles and advance the scale-up, design, and performance of TRISO fuel particle spouted bed coaters. The overall objectives of the proposed work are

Pyro-electrochemical reprocessing of irradiated MOX fast reactor fuel, testing of the reprocessing process with direct MOX fuel production

Energy Technology Data Exchange (ETDEWEB)

Kormilitzyn, M.V.; Vavilov, S.K.; Bychkov, A.V.; Skiba, O.V.; Chistyakov, V.M.; Tselichshev, I.V

2000-07-01

One of the advanced technologies for fast reactor fuel recycle is pyro-electrochemical molten salt technology. In 1998 we began to study the next phase of the irradiated oxide fuel reprocessing new process MOX {yields} MOX. This process involves the following steps: - Dissolution of irradiated fuel in molten alkaline metal chlorides, - Purification of melt from fission products that are co-deposited with uranium and plutonium oxides, - Electrochemical co-deposition of uranium and plutonium oxides under the controlled cathode potential, - Production of granulated MOX (crushing,salt separation and sizing), and - Purification of melt from fission products by phosphate precipitation. In 1998 a series of experiments were prepared and carried out in order to validate this process. It was shown that the proposed reprocessing flowsheet of irradiated MOX fuel verified the feasibility of its decontamination from most of its fission products (rare earths, cesium) and minor-actinides (americium, curium)

Pyro-electrochemical reprocessing of irradiated MOX fast reactor fuel, testing of the reprocessing process with direct MOX fuel production

International Nuclear Information System (INIS)

Kormilitzyn, M.V.; Vavilov, S.K.; Bychkov, A.V.; Skiba, O.V.; Chistyakov, V.M.; Tselichshev, I.V.

2000-01-01

One of the advanced technologies for fast reactor fuel recycle is pyro-electrochemical molten salt technology. In 1998 we began to study the next phase of the irradiated oxide fuel reprocessing new process MOX → MOX. This process involves the following steps: - Dissolution of irradiated fuel in molten alkaline metal chlorides, - Purification of melt from fission products that are co-deposited with uranium and plutonium oxides, - Electrochemical co-deposition of uranium and plutonium oxides under the controlled cathode potential, - Production of granulated MOX (crushing,salt separation and sizing), and - Purification of melt from fission products by phosphate precipitation. In 1998 a series of experiments were prepared and carried out in order to validate this process. It was shown that the proposed reprocessing flowsheet of irradiated MOX fuel verified the feasibility of its decontamination from most of its fission products (rare earths, cesium) and minor-actinides (americium, curium)

Advanced fuel technologies at General Atomics

International Nuclear Information System (INIS)

Back, Christina A.

2013-01-01

General Atomics (GA) has made significant contributions since its founding in the 1950's to develop nuclear power for peaceful means. With the conception and construction of the TRIGA reactors and research on TRISO particles, GA has long recognised the importance of 'accident-tolerant' materials. Before the accident at Fukushima Daiichi, GA had already initiated work on silicon carbide (SiC) and SiC-related technologies for application in nuclear reactors. At that time, the work was initiated in support of the GA advanced gas-cooled fast reactor concept called the Energy Multiplier Module, EM2. This work continues, however, the reasons that make SiC materials attractive for fast reactor concepts also make them attractive for advanced light water reactors. These include superior performance over zircaloy for high-temperature strength, especially above 1500 deg. C, and significantly reduced hydrogen production in accident scenarios. The current focus on 'accident-tolerant' components is to develop cladding made of silicon carbide fiber and silicon carbide matrix, SiC-SiC composites. The goal for this work is to produce a cladding that provides strength and impermeability to meet reactor performance and safety requirements. To date, GA has examined the trade-offs between processing time and infiltration uniformity to reduce fabrication time, fabricated cylindrical prototypes, and refined material properties for fracture toughness, impermeability, and thermal conductivity. Generally, the GA programme is developing innovative fuel elements that employ both high density uranium-bearing fuels that enable longer lifetime with higher burn-up, and claddings that are more resistant to neutron damage. In addition to fabrication, significant effort is devoted to measuring the critical parameters, such as thermal conductivity, mechanical strength and component performance at reactor-relevant operational conditions, using a mix of commercial equipment

CANDU advanced fuel cycles

International Nuclear Information System (INIS)

Slater, J.B.

1986-03-01

This report is based on informal lectures and presentations made on CANDU Advanced Fuel Cycles over the past year or so, and discusses the future role of CANDU in the changing environment for the Canadian and international nuclear power industry. The changing perspectives of the past decade lead to the conclusion that a significant future market for a CANDU advanced thermal reactor will exist for many decades. Such a reactor could operate in a stand-alone strategy or integrate with a mixed CANDU-LWR or CANDU-FBR strategy. The consistent design focus of CANDU on enhanced efficiency of resource utilization combined with a simple technology to achieve economic targets, will provide sufficient flexibility to maintain CANDU as a viable power producer for both the medium- and long-term future

Conceptual framework for using 'Best Estimate plus Uncertainty' as a basis for licensing activities for fuels developed for an advanced reactor

International Nuclear Information System (INIS)

McClure, P.; Unal, C.; Boyack, B.

2010-01-01

Closing the fuel cycle is one of the major technical challenges to expanding the use of nuclear energy to meet the world's need for benign, environmentally safe electrical power. 'Closing the fuel cycle ' means getting the maximum amount of energy possible out of uranium fuel while minimizing the amount of high-level waste that must be stored. The U.S. Dept. of Energy's Fuel Cycle Research and Development (FCRD) program is investigating the recycling of transuranic isotopes contained in spent nuclear fuel. Recycling minimizes the amount of high-level waste that would require storage in repositories. Developing new fuels and the advanced reactors that burn them is a long process typically spanning two decades from concept to final licensing. A unique challenge to meeting the FCRD objectives in this area is the fact that the experimental database is incomplete. Thus, using a traditional, heavily empirical approach to develop and qualify fuels for an advanced reactor plant will be very challenging. To address this concern, FCRD has launched an advanced modeling and simulation (M and S) approach to revolutionize fuel development and advanced reactor design. This new approach depends on transferring recent advances in the computational sciences and computer technologies into the development of these program elements. The licensing process that historically has been used by the U.S. Nuclear Regulatory Commission (NRC) for fuels qualification is based on using a large body of experimental work to qualify and license a new fuel. If an M and S approach with more directed experimentation is to be considered as an alternative approach for licensing, a framework needs to be developed early in the process. Using M and S with limited experiments as a basis for demonstrating that a design can meet NRC requirements is not new and has precedence in the NRC. The method is generically referred to as a 'Best Estimate plus Uncertainty' (BE+U) approach because the goal of the

Experience related to the safety of advanced LMFBR fuel elements

International Nuclear Information System (INIS)

Kerrisk, J.F.

1975-07-01

Experiments and experience relative to the safety of advanced fuel elements for the liquid metal fast breeder reactor are reviewed. The design and operating parameters and some of the unique features of advanced fuel elements are discussed breifly. Transient and steady state overpower operation and loss of sodium bond tests and experience are discussed in detail. Areas where information is lacking are also mentioned

Shielding analysis of the advanced voloxidation process

Energy Technology Data Exchange (ETDEWEB)

Park, Chang Je; Park, J. J.; Lee, J. W.; Shin, J. M.; Park, G. I.; Song, K. C

2008-09-15

This report deals describes how much a shielding benefit can be obtained by the Advanced Voloxidation process. The calculation was performed with the MCNPX code and a simple problem was modeled with a spent fuel source which was surrounded by a concrete wall. The source terms were estimated with the ORIGEN-ARP code and the gamma spectrum and the neutron spectrum were also obtained. The thickness of the concrete wall was estimated before and after the voloxidation process. From the results, the gamma spectrum after the voloxidation process was estimated as a 67% reduction compared with that of before the voloxidation process due to the removal of several gamma emission elements such as cesium and rubidium. The MCNPX calculations provided that the thickness of the general concrete wall could be reduced by 12% after the voloxidation process. And the heavy concrete wall provided a 28% reduction in the shielding of the source term after the voloxidation process. This can be explained in that there lots of gamma emission isotopes still exist after the advanced voloxidation process such as Pu-241, Y-90, and Sr-90 which are independent of the voloxidation process.

Impact of advanced fuel cycle options on waste management policies

International Nuclear Information System (INIS)

Gordelier, Stan; Cavedon, Jean-Marc

2006-01-01

OECD/NEA has performed a study on the impact of advanced fuel cycle options on waste management policies with 33 experts from 12 member countries, 1 non-member country and 2 international organizations. The study extends a series of previous ones on partitioning and transmutation (P and T) issues, focusing on the performance assessments for repositories of high-level waste (HLW) arising from advanced fuel cycles. This study covers a broader spectrum than previous studies, from present industrial practice to fully closed cycles via partially closed cycles (in terms of transuranic elements); 9 fuel cycle schemes and 4 variants. Elements of fuel cycles are considered primarily as sources of waste, the internal mass flows of each scheme being kept for the sake of mass conservation. The compositions, activities and heat loads of all waste flows are also tracked. Their impact is finally assessed on the waste repository concepts. The study result confirms the findings from the previous NEA studies on P and T on maximal reduction of the waste source term and maximal use of uranium resources. In advanced fuel cycle schemes the activity of the waste is reduced by burning first plutonium and then minor actinides and also the uranium consumption is reduced, as the fraction of fast reactors in the park is increased to 100%. The result of the repository performance assessments, analysing the effect of different HLW isotopic composition on repository performance and on repository capacity, shows that the maximum dose released to biosphere at any time in normal conditions remains, for all schemes and for all the repository concepts examined, well below accepted radiation protection thresholds. The major impact is on the detailed concept of the repositories, through heat load and waste volume. Advanced fuel cycles could allow a repository to cover waste produced from 5 to 20 times more electricity generation than PWR once-through cycle. Given the flexibility of the advanced fuel

Report of 5th new nuclear fuel research meeting, Yayoi Research Group. Trend of advanced basic research in nuclear fuel technical development

International Nuclear Information System (INIS)

1994-03-01

Theme of this meeting is 'Trend of advanced basic research in nuclear fuel technical development', and it was attempted to balance both sides of the basic research and the development. At the meeting, lectures were given on the chemical form of FPs in oxide fuel pins, the absorption of hydrogen of fuel cladding tubes, the application of hydride fuel to thorium cycle, the thermal properties of fuel cladding tubes, the preparation of NpN and heat conductivity, the high temperature chemical reprocessing of nitride fuel, the research on the annihilation treatment of minor actinide in fast reactors, the separation of TRU by dry process and the annihilation using a metallic fuel FBR. In this report, the summaries of the lectures are collected, and also the program of the meeting and the list of attendants are shown. (K.I.)

High Efficiency Advanced Lightweight Fuel Cell (HEAL-FC), Phase I

Data.gov (United States)

National Aeronautics and Space Administration — Infinity's High Efficiency Advanced Lightweight Fuel Cell (HEAL FC) is an improved version of its current fuel cell technology developed for space applications. The...

Corrosion of spent Advanced Test Reactor fuel

International Nuclear Information System (INIS)

Lundberg, L.B.; Croson, M.L.

1994-01-01

The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented

Advances in uranium enrichment processes

International Nuclear Information System (INIS)

Rae, H.K.; Melvin, J.G.; Slater, J.B.

1986-05-01

Advances in gas centrifuges and development of the atomic vapour laser isotope separation process promise substantial reductions in the cost of enriched uranium. The resulting reduction in LWR fuel costs could seriously erode the economic advantage of CANDU, and in combination with LWR design improvements, shortened construction times and increased operational reliability could allow the LWR to overtake CANDU. CANDU's traditional advantages of neutron economy and high reliability may no longer be sufficient - this is the challenge. The responses include: combining neutron economy and dollar economy by optimizing CANDU for slightly enriched uranium fuel; developing cost-reducing improvements in design, manufacture and construction; and reducing the cost of heavy water. Technology is a renewable resource which must be continually applied to a product for it to remain competitive in the decades to come. Such innovation is a prerequisite to Canada increasing her share of the international market for nuclear power stations. The higher burn-up achievable with enriched fuel in CANDU can reduce the fuel cycle costs by 20 to 40 percent for a likely range of costs for yellowcake and separative work. Alternatively, some of the benefits of a higher fissile content can take the form of a cheaper reactor core containing fewer fuel channels and less heavy water, and needing only a single fuelling machine. An opportunity that is linked to this need to introduce an enriched uranium fuel cycle into CANDU is to build an enrichment business in Canada. This could offer greater value added to our uranium exports, security of supply for enriched CANDUs, technological growth in Canada and new employment opportunities. AECL has a study in progress to define this opportunity

Microstructure characterizaton of advanced oxide fuel

International Nuclear Information System (INIS)

Rasmussen, D.E.; Gerber, E.W.; McCord, R.B.

1977-01-01

Preirradiation porosity, grain size, and microcomposition characteristics are presented for selected advanced oxide (PuO 2 -UO 2 ) LMFBR developmental fuels fabricated for irradiation testing in EBR-II. Quantitative microscopy, electron microprobe analysis, and a recently developed quantitative autoradiographic technique are utilized to relate microstructure characteristics to fabrication parameters

Advanced Research Reactor Fuel Development

Energy Technology Data Exchange (ETDEWEB)

Kim, C. K.; Park, H. D.; Kim, K. H. (and others)

2006-04-15

RERTR program for non-proliferation has propelled to develop high-density U-Mo dispersion fuels, reprocessable and available as nuclear fuel for high performance research reactors in the world. As the centrifugal atomization technology, invented in KAERI, is optimum to fabricate high-density U-Mo fuel powders, it has a great possibility to be applied in commercialization if the atomized fuel shows an acceptable in-reactor performance in irradiation test for qualification. In addition, if rod-type U-Mo dispersion fuel is developed for qualification, it is a great possibility to export the HANARO technology and the U-Mo dispersion fuel to the research reactors supplied in foreign countries in future. In this project, reprocessable rod-type U-Mo test fuel was fabricated, and irradiated in HANARO. New U-Mo fuel to suppress the interaction between U-Mo and Al matrix was designed and evaluated for in-reactor irradiation test. The fabrication process of new U-Mo fuel developed, and the irradiation test fuel was fabricated. In-reactor irradiation data for practical use of U-Mo fuel was collected and evaluated. Application plan of atomized U-Mo powder to the commercialization of U-Mo fuel was investigated.

Development of proto-type advanced leaked fuel rod detection system

International Nuclear Information System (INIS)

Kang, Kyung Chul; Cho, Seong Won; Jeon, Jae Hyuk; Jeong, Jae Cheon; Kim, Min

1996-02-01

The fuel inspection equipment using ultrasonic signal has been developed its design and configuration in order to get inspection results more accurate and easier than the previous ones. In this task, the system functions are advanced by adopting of state of the art technologies in the field of digital servo control and signal processing. By the above endeavors, the total performance are improved and made to handle easily. 61 tabs., 31 figs., 3 ills., 9 refs. (Author)

Outline of an experimental apparatus for the study on the advanced voloxidation process

International Nuclear Information System (INIS)

Uchiyama, Gunzo; Sugikawa, Susumu; Maeda, Mitsuru; Tsujino, Takeshi; Torikai, Seishi; Kitamura, Masafumi; Yamazaki, Kazunobu.

1990-02-01

The experimental apparatus (VULCAN, the capacity; 2 kg-UO 2 /batch) was constructed to study on the advanced voloxidation process, which was proposed to reduce amount of tritium released from fuel reprocessing facilities. Using this equipment, a process study was conducted on behaviors of oxidation-reduction of simulated fuels and of release of tritium, and on confinement function of rotary seal of the reactor. An outline of the experimental apparatus is described. (author)

Advanced analysis technology for MOX fuel

International Nuclear Information System (INIS)

Hiyama, T.; Kamimura, K.

1997-01-01

PNC has developed MOX fuels for advanced thermal reactor (ATR) and fast breeder reactor (FBR). The MOX samples have been chemically analysed to characterize the MOX fuel for JOYO, MONJU, FUGEN and so on. The analysis of the MOX samples in glove box has required complicated and highly skilled operations. Therefore, for quality control analysis of the MOX fuel in a fabrication plant, simple, rapid and accurate analysis methods are necessary. To solve the above problems instrumental analysis and techniques were developed. This paper describes some of the recent developments in PNC. 2. Outline of recently developed analysis methods by PNC. 2.1 Determination of oxygen to metal atomic ratio (O/M) in MOX by non-dispersive infrared spectrophotometry after inert gas fusion. 7 refs, 9 figs, 4 tabs

Development of the advanced PHWR technology -Verification tests for CANDU advanced fuel-

Energy Technology Data Exchange (ETDEWEB)

Jung, Jang Hwan; Suk, Hoh Chun; Jung, Moon Kee; Oh, Duk Joo; Park, Joo Hwan; Shim, Kee Sub; Jang, Suk Kyoo; Jung, Heung Joon; Park, Jin Suk; Jung, Seung Hoh; Jun, Ji Soo; Lee, Yung Wook; Jung, Chang Joon; Byun, Taek Sang; Park, Kwang Suk; Kim, Bok Deuk; Min, Kyung Hoh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

This is the `94 annual report of the CANDU advanced fuel verification test project. This report describes the out-of pile hydraulic tests at CANDU-hot test loop for verification of CANFLEX fuel bundle. It is also describes the reactor thermal-hydraulic analysis for thermal margin and flow stability. The contents in this report are as follows; (1) Out-of pile hydraulic tests for verification of CANFLEX fuel bundle. (a) Pressure drop tests at reactor operation condition (b) Strength test during reload at static condition (c) Impact test during reload at impact load condition (d) Endurance test for verification of fuel integrity during life time (2) Reactor thermal-hydraulic analysis with CANFLEX fuel bundle. (a) Critical channel power sensitivity analysis (b) CANDU-6 channel flow analysis (c) Flow instability analysis. 61 figs, 29 tabs, 21 refs. (Author).

Development of the advanced PHWR technology -Verification tests for CANDU advanced fuel-

International Nuclear Information System (INIS)

Jung, Jang Hwan; Suk, Hoh Chun; Jung, Moon Kee; Oh, Duk Joo; Park, Joo Hwan; Shim, Kee Sub; Jang, Suk Kyoo; Jung, Heung Joon; Park, Jin Suk; Jung, Seung Hoh; Jun, Ji Soo; Lee, Yung Wook; Jung, Chang Joon; Byun, Taek Sang; Park, Kwang Suk; Kim, Bok Deuk; Min, Kyung Hoh

1995-07-01

This is the '94 annual report of the CANDU advanced fuel verification test project. This report describes the out-of pile hydraulic tests at CANDU-hot test loop for verification of CANFLEX fuel bundle. It is also describes the reactor thermal-hydraulic analysis for thermal margin and flow stability. The contents in this report are as follows; (1) Out-of pile hydraulic tests for verification of CANFLEX fuel bundle. (a) Pressure drop tests at reactor operation condition (b) Strength test during reload at static condition (c) Impact test during reload at impact load condition (d) Endurance test for verification of fuel integrity during life time (2) Reactor thermal-hydraulic analysis with CANFLEX fuel bundle. (a) Critical channel power sensitivity analysis (b) CANDU-6 channel flow analysis (c) Flow instability analysis. 61 figs, 29 tabs, 21 refs. (Author)

Advanced diesel electronic fuel injection and turbocharging

Science.gov (United States)

Beck, N. J.; Barkhimer, R. L.; Steinmeyer, D. C.; Kelly, J. E.

1993-12-01

The program investigated advanced diesel air charging and fuel injection systems to improve specific power, fuel economy, noise, exhaust emissions, and cold startability. The techniques explored included variable fuel injection rate shaping, variable injection timing, full-authority electronic engine control, turbo-compound cooling, regenerative air circulation as a cold start aid, and variable geometry turbocharging. A Servojet electronic fuel injection system was designed and manufactured for the Cummins VTA-903 engine. A special Servojet twin turbocharger exhaust system was also installed. A series of high speed combustion flame photos was taken using the single cylinder optical engine at Michigan Technological University. Various fuel injection rate shapes and nozzle configurations were evaluated. Single-cylinder bench tests were performed to evaluate regenerative inlet air heating techniques as an aid to cold starting. An exhaust-driven axial cooling air fan was manufactured and tested on the VTA-903 engine.

Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

International Nuclear Information System (INIS)

Lu, Hongbing; Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

2014-01-01

Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

Energy Technology Data Exchange (ETDEWEB)

Lu, Hongbing [Univ. of Texas, Austin, TX (United States); Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

2014-01-09

Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

Assessment of Research Needs for Advanced Fuel Cells

Energy Technology Data Exchange (ETDEWEB)

Penner, S.S.

1985-11-01

The DOE Advanced Fuel Cell Working Group (AFCWG) was formed and asked to perform a scientific evaluation of the current status of fuel cells, with emphasis on identification of long-range research that may have a significant impact on the practical utilization of fuel cells in a variety of applications. The AFCWG held six meetings at locations throughout the country where fuel cell research and development are in progress, for presentations by experts on the status of fuel cell research and development efforts, as well as for inputs on research needs. Subsequent discussions by the AFCWG have resulted in the identification of priority research areas that should be explored over the long term in order to advance the design and performance of fuel cells of all types. Surveys describing the salient features of individual fuel cell types are presented in Chapters 2 to 6 and include elaborations of long-term research needs relating to the expeditious introduction of improved fuel cells. The Introduction and the Summary (Chapter 1) were prepared by AFCWG. They were repeatedly revised in response to comments and criticism. The present version represents the closest approach to a consensus that we were able to reach, which should not be interpreted to mean that each member of AFCWG endorses every statement and every unexpressed deletion. The Introduction and Summary always represent a majority view and, occasionally, a unanimous judgment. Chapters 2 to 6 provide background information and carry the names of identified authors. The identified authors of Chapters 2 to 6, rather than AFCWG as a whole, bear full responsibility for the scientific and technical contents of these chapters.

U.S. Department of Energy & Nuclear Regulatory Commission Advanced Fuel Cycle Research & Development Seminar Series FY 2007 & 2008

Energy Technology Data Exchange (ETDEWEB)

Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States)

2008-08-01

In fiscal year 2007, the Advanced Burner Reactor project initiated an educational seminar series for the Department of Energy (DOE) and Nuclear Regulatory Commission (NRC) personnel on various aspects of fast reactor fuel cycle closure technologies. This important work was initiated to inform DOE and NRC personnel on initial details of sodium-cooled fast reactor, separations, waste form, and safeguard technologies being considered for the Advanced Fuel Cycle Research and Development program, and to learn the important lesson from the licensing process for the Clinch River Breeder Reactor Plant that educating the NRC staff early in the regulatory process is very important and critical to a project success.

Safeguardability assessment on pilot-scale advanced spent fuel conditioning facility

International Nuclear Information System (INIS)

Lee, S.Y.; Li, T.K.; Pickett, S.E.; Miller, M.C.; Ko, W.I.; Kim, H.D.

2006-01-01

Full text: In South Korea, approximately 6,000 metric tons of spent nuclear fuel from commercial reactor operation has been accumulated with the expectation of more than 30,000 metric tons, three times the present storage capacity, by the end of 2040. To resolve these challenges in spent fuel management, the Korea Atomic Energy Research Institute (KAERI) has been developing a dry reprocessing technology called Advanced Spent Fuel Conditioning Process (ACP). This is an electrometallurgical treatment technique to convert oxide-type spent fuel into a metallic form, and the electrolytic reduction (ER) technology developed recently is known as a more efficient concept for spent fuel conditioning. The goal of the ACP study is to recover more than 99% of the actinide elements into a metallic form with minimizing the volume and heat load of spent fuel. The significant reduction of the volume and heat load of spent fuel is expected to lighten the burden of final disposal in terms of disposal size, safety, and economics. In the framework of R and D collaboration for the ACP safeguards, a joint study on the safeguardability of the ACP technology has been performed by the Los Alamos National Laboratory (LANL) and KAERI. The purpose of this study is to address the safeguardability of the ACP technology, through analysis of material flow and development of a proper safeguards system that meet IAEA's comprehensive safeguards objective. The sub-processes and material flow of the pilot-scale ACP facility were analyzed, and subsequently the relevant material balance area (MBA) and key measurement point (KMP) were designed for material accounting. The uncertainties in material accounting were also estimated with international target values, and design requirements for the material accounting systems were derived

Physics characteristics of CANDU cores with advanced fuel cycles

International Nuclear Information System (INIS)

Garvey, P.M.

1985-01-01

The current generation of CANDU reactors, of which some 20 GWE are either in operations or under construction worldwide, have been designed specifically for the natural uranium fuel cycle. The CANDU concept, due to its D 2 O coolant and moderator, on-power refuelling and low absorption structural materials, makes the most effective utilization of mined uranium of all currently commercialized reactors. An economic fuel cycle cost is also achieved through the use of natural uranium and a simple fuel bundle design. Total unit energy costs are achieved that allow this reactor concept to effectively compete with other reactor types and other forms of energy production. There are, however, other fuel cycles that could be introduced into this reactor type. These include the slightly enriched uranium fuel cycle, fuel cycles in which plutonium is recycled with uranium, and the thorium cycle in which U-233 is recycled. There is also a special range of fuel cycles that could utilize the spent fuel from LWR's. Two specific variants are a fuel cycle that only utilizes the spent uranium, and a fuel cycle in which both the uranium and plutonium are recycled into a CANDU. For the main part these fuel cycles are characterized by a higher initial enrichment, and hence discharge burnup, than the natural uranium cycle. For these fuel cycles the main design features of both the reactor and fuel bundle would be retained. Recently a detailed study of the use in a CANDU of mixed plutonium and uranium oxide fuel from an LWR has been undertaken by AECL. This study illustrates many of the generic technical issues associated with the use of Advanced Fuel Cycles. This paper will report the main findings of this evaluation, including the power distribution in the reactor and fuel bundle, the choice of fuel management scheme, and the impact on the control and safety characteristics of the reactor. These studies have not identified any aspects that significantly impact upon the introduction of

Design study on advanced nuclear fuel recycle system. Conceptual design study of recycle system using molten salt

International Nuclear Information System (INIS)

Kasai, Y.; Kakehi, I.; Moro, T.; Higashi, T.; Tobe, K.; Kawamura, F.; Yonezawa, S.; Yoshiuji, T.

1998-10-01

Advanced recycle system engineering group of OEC (Oarai Engineering Center) has being carried out a design study of the advanced nuclear fuel recycle system using molten salt (electro-metallurgical process). This system is aiming for improvements of fuel cycle economy and reduction of environmental burden (MA recycles, Minimum of radioactive waste disposal), and also improvement of safety and nuclear non-proliferation. This report describes results of the design study that has been continued since December 1996. (1) A design concept of the advanced nuclear fuel recycle system, that is a module type recycles system of pyrochemical reprocessing and fuel re-fabrication was studied. The module system has advantage in balance of Pu recycle where modules are constructed in coincidence with the construction plan of nuclear power plants, and also has flexibility for technology progress. A demonstration system, minimum size of the above module, was studies. This system has capacity of 10 tHM/y and is able to demonstrate recycle technology of MOX fuel, metal fuel and nitride fuel. (2) Each process of the system, which are pyrochemical electrorefining system, cathode processor, de-cladding system, waste disposal system, etc., were studied. In this study, capacity of an electrorefiner was discussed, and vitrification experiment of molten salt using lead-boric acid glass was conducted. (3) A hot cell system and material handling system of the demonstration system was studied. A robot driven by linear motor was studied for the handling system, and an arrangement plan of the cell system was made. Criticality analysis in the cell system and investigation of material accountancy system of the recycle plant were also made. This design study will be continued in coincidence with design study of reactor and fuel, aiming to establish the concept of FBR recycle system. (author)

Potential role of advanced fuels in inertial confinement fusion

International Nuclear Information System (INIS)

Miley, G.

1981-01-01

The potential importance of developing advanced (non D-T) fuel pellets for inertial confinement is discussed. Reduced radioactivity due to low tritium involvement and less neutron activation, improved blanket flexibility with the removal of tritium breeding requirements, and improved mating of the output energy spectrum with non-electrical applications such as synthetic fuel production could lead to technical advantages and earlier public acceptance. As a possible first step to advanced-fuel pellets, the A-FLINT concept of a D-T core ignited, deuterium pellet is proposed which would offer tritium self-sufficiency. A design is described that uses 0.1-MJ internal energy in a rhoR1--7 gm/cm2'' compressed pellet, giving a tritium breeding ratio of 1--1.0 and an internal pellet gain of 1--700

Irradiation performance of (Th,U)O2 fuel designed for advanced cycle applications

International Nuclear Information System (INIS)

Hastings, I.J.; Celli, A.; Onofrei, M.; Swanson, M.L.

1982-06-01

Our reference fabrication route for Advanced Cycle thoria-based fuel is conventional in that it produces cold-pressed and sintered pellets. However, we are also evaluating alternative fuels which offer the potential for simpler fabrication in a remote facility, and in some cases improved high burnup performance. These alternatives are impregnated, spherepac, and extruded thoria-based fuels. Spherepac fuel has been irradiated at a linear power of 50-60 kW/m to about 180 MW.h/kg H.E. There have been unexplained defects in fuel with both free-standing and collapsible cladding. Impegnated fuel has operated to 650 MW.h/kg H.E. at 50-60 kW/m. An experiment examining fuel from the sol-gel extrusion process has reached 450 Mw.h/kg H.E. at a maximum linear power of 60 kW/m. The latter two experiments have operated without defects and with fission gas release less than that for UO 2 under identical conditions. The extruded fuel has a pellet geometry similar to that for conventional fuel and is AECL's first practical demonstration of thoria-based fuel with the fissile component distributed homogeneously on an atomic scale

Fuel fragmentation model advances using TEXAS-V

Energy Technology Data Exchange (ETDEWEB)

Corradini, M.L.; El-Beshbeeshy, M.; Nilsuwankowsit, S.; Tang, J. [Wisconsin Univ., Madison, WI (United States). Dept. of Nuclear Engineering and Engineering Physics

1998-01-01

Because an energetic fuel-coolant interaction may be a safety hazard, experiments are being conducted to investigate the fuel-coolant mixing/quenching process (FARO) as well as the energetics of vapor explosion propagation for high temperature fuel melt simulants (KROTOS, WFCI, ZrEX). In both types of experiments, the dynamic breakup of the fuel is one of the key aspects that must be fundamentally understood to better estimate the magnitude of the mixing/quenching process or the explosion energetics. To aid our understanding the TEXAS fuel-coolant interaction computer model has been developed and is being used to analyze these experiments. Recently, the models for dynamic fuel fragmentation during the mixing and explosion phases of the FCI have been improved by further insights into these processes. The purpose of this paper is to describe these enhancements and to demonstrate their improvements by analysis of particular JRC FCI data. (author)

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.

The fuel to clad heat transfer coefficient in advanced MX-type fuel pins

International Nuclear Information System (INIS)

Caligara, F.; Campana, M.; Mandler, R.; Blank, H.

1979-01-01

Advanced fuels (mixed carbides, nitrides and carbonitrides) are characterised by a high thermal conductivity compared to that of oxide fuels (5 times greater) and their behaviour under irradiation (amount of swelling, fracture behaviour, restructuring) is far more sensitive to the design parameters and to the operating temperature than that of oxide fuels. The use of advanced fuels is therefore conditioned by the possibility of mastering the above phenomena, and the full exploitation of their favorable neutron characteristics depends upon a good understanding of the mutual relationships of the various parameters, which eventually affect the mechanical stability of the pin. By far the most important parameter is the radial temperature profile which controls the swelling of the fuel and the build-up of stress fields within the pin. Since the rate of fission gas swelling of these fuels is relatively large, a sufficient amount of free space has to be provided within the pin. This space originally appears as fabrication porosity and as fuel-to-clad clearance. Due to the large initial gap width and to the high fuel thermal conductivity, the range of the fuel operating temperatures is mainly determined by the fuel-to-clad heat transfer coefficient h, whose correct determination becomes one of the central points in modelling. During the many years of modelling activity in the field of oxide fuels, several theoretical models have been developed to calculate h, and a large amount of experimental data has been produced for the empirical adjustment of the parameters involved, so that the situation may be regarded as rather satisfactory. The analysis lead to the following conclusions. A quantitative comparison of experimental h-values with existing models for h requires rather sophisticated instrumented irradiation capsules, which permit the measurement of mechanical data (concerning fuel and clad) together with heat rating and temperatures. More and better well

Technology readiness level (TRL) assessment of cladding alloys for advanced nuclear fuels

International Nuclear Information System (INIS)

Shepherd, Daniel

2015-01-01

Reliable fuel claddings are essential for the safe, sustainable and economic operation of nuclear stations. This paper presents a worldwide TRL assessment of advanced claddings for Gen III and IV reactors following an extensive literature review. Claddings include austenitic, ferritic/martensitic (F/M), reduced activation (RA) and oxide dispersion strengthened (ODS) steels as well as advanced iron-based alloys (Kanthal alloys). Also assessed are alloys of zirconium, nickel (including Hastelloy R ), titanium, chromium, vanadium and refractory metals (Nb, Mo, Ta and W). Comparison is made with Cf/C and SiCf/SiC composites, MAX phase ceramics, cermets and TRISO fuel particle coatings. The results show in general that the higher the maximum operating temperature of the cladding, the lower the TRL. Advanced claddings were found to have lower TRLs than the corresponding fuel materials, and therefore may be the limiting factor in the deployment of advanced fuels and even possibly the entire reactor in the case of Gen IV. (authors)

Development of advanced mixed oxide fuels for plutonium management

International Nuclear Information System (INIS)

Eaton, S.; Beard, C.; Buksa, J.; Butt, D.; Chidester, K.; Havrilla, G.; Ramsey, K.

1997-01-01

A number of advanced Mixed Oxide (MOX) fuel forms are currently being investigated at Los Alamos National Laboratory that have the potential to be effective plutonium management tools. Evolutionary Mixed Oxide (EMOX) fuel is a slight perturbation on standard MOX fuel, but achieves greater plutonium destruction rates by employing a fractional nonfertile component. A pure nonfertile fuel is also being studied. Initial calculations show that the fuel can be utilized in existing light water reactors and tailored to address different plutonium management goals (i.e., stabilization or reduction of plutonium inventories residing in spent nuclear fuel). In parallel, experiments are being performed to determine the feasibility of fabrication of such fuels. Initial EMOX pellets have successfully been fabricated using weapons-grade plutonium. (author)

Development of advanced mixed oxide fuels for plutonium management

International Nuclear Information System (INIS)

Eaton, S.; Beard, C.; Buksa, J.; Butt, D.; Chidester, K.; Havrilla, G.; Ramsey, K.

1997-06-01

A number of advanced Mixed Oxide (MOX) fuel forms are currently being investigated at Los Alamos National Laboratory that have the potential to be effective plutonium management tools. Evolutionary Mixed Oxide (EMOX) fuel is a slight perturbation on standard MOX fuel, but achieves greater plutonium destruction rates by employing a fractional nonfertile component. A pure nonfertile fuel is also being studied. Initial calculations show that the fuel can be utilized in existing light water reactors and tailored to address different plutonium management goals (i.e., stabilization or reduction of plutonium inventories residing in spent nuclear fuel). In parallel, experiments are being performed to determine the feasibility of fabrication of such fuels. Initial EMOX pellets have successfully been fabricated using weapons-grade plutonium

FCRD Advanced Reactor (Transmutation) Fuels Handbook

Energy Technology Data Exchange (ETDEWEB)

Janney, Dawn Elizabeth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Papesch, Cynthia Ann [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. One important approach to transmutation (currently being pursued by the DOE Fuel Cycle Research & Development Advanced Fuels Campaign) involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. U-Pu-Zr alloys are well suited for electrolytic refining, which leads to incorporation rare-earth fission products such as La, Ce, Pr, and Nd. It is, therefore, important to understand not only the properties of U-Pu-Zr alloys but also those of U-Pu-Zr alloys with concentrations of minor actinides (Np, Am) and rare-earth elements (La, Ce, Pr, and Nd) similar to those in reprocessed fuel. In addition to requiring extensive safety precautions, alloys containing U, Pu, and minor actinides (Np and Am) are difficult to study for numerous reasons, including their complex phase transformations, characteristically sluggish phasetransformation kinetics, tendency to produce experimental results that vary depending on the histories of individual samples, rapid oxidation, and sensitivity to contaminants such as oxygen in concentrations below a hundred parts per million. Although less toxic, rare-earth elements such as La, Ce, Pr, and Nd are also difficult to study for similar reasons. Many of the experimental measurements were made before 1980, and the level of documentation for experimental methods and results varies widely. It is, therefore, not surprising that little is known with certainty about U-Pu-Zr alloys, particularly those that also contain minor actinides and rare-earth elements. General acceptance of results commonly indicates that there is only a single measurement for a particular property. This handbook summarizes currently available information about U, Pu, Zr, Np, Am, La, Ce, Pr, and Nd and

A contingency safe, responsible, economic, increased capacity spent nuclear fuel (SNF) advance fuel cycle

International Nuclear Information System (INIS)

Levy, S.

2008-01-01

The purpose of this paper is to have an Advanced Light Water (LWR) fuel cycle and an associated development program to provide a contingency plan to the current DOE effort to license once-through spent Light Water Reactor (LWR) fuel for disposition at Yucca Mountain (YM). The intent is to fully support the forthcoming June 2008 DOE submittal to the Nuclear Regulatory Commission (NRC) based upon the latest DOE draft DOE/EIS-0250F-SID dated October 2007 which shows that the latest DOE YM doses would readily satisfy the anticipated NRC and Environmental Protection Agency (EP) standards. The proposed Advance Fuel Cycle can offer potential resolution of obstacles that might arise during the NRC review and, particularly, during the final hearings process to be held in Nevada. Another reason for the proposed concept is that a substantial capacity growth of the YM repository will be necessary to accommodate the SNF of Advance Light Water Reactors (ALWRs) currently under consideration for United States (U.S.) electricity production (1) and the results of the recently issued study by the Electric Power Research Institute (EPRI) to reduce CO 2 emissions (2). That study predicts that by 2030 U.S. nuclear power generation would grow by 64 Gigawatt electrical (GWe) and account for 25.5 percent of the overall U.S. electrical generation. The current annual SNF once-through fuel cycle accumulation would rise from 2000-2100 MT (Metric Tons) to about 3480 MT in 2030 and the total SNF inventory, would reach nearly 500,000 MT by 2100 if U. S. nuclear power continues to grow at 1.1 percent per year after 2030. That last projection does not account for any SNF reduction due to increased fuel burnup or any increased capacity needed 'to establish supply Global Nuclear Energy Partnership (GNEP,) arrangements among nations to provide nuclear fuel and taking back spent fuel for recycling without spreading enrichment and reprocessing technologies' (3). The anticipated capacity of 120 MT

Use of a hot sheath Tormac for advance fuels

International Nuclear Information System (INIS)

Levine, M.A.

1977-01-01

The use of hot electrons in a Tormac sheath is predicted to improve stability and increase ntau by an order of magnitude. An effective ntau for energy containment is derived and system parameters for several advance fuels are shown. In none of the advance fuels cases considered is a reactor with fields greater than 10 Wb or major plasma radius of more than 3 m required for ignition. Minimum systems have power output of under 100 MW thermal. System parameters for a hot sheath Tormac have a wide latitude. Sizes, magnetic fields, operating temperatures can be chosen to optimize engineering and economic considerations

Progress on matrix SiC processing and properties for fully ceramic microencapsulated fuel form

International Nuclear Information System (INIS)

Terrani, K.A.; Kiggans, J.O.; Silva, C.M.; Shih, C.; Katoh, Y.; Snead, L.L.

2015-01-01

The consolidation mechanism and resulting properties of the silicon carbide (SiC) matrix of fully ceramic microencapsulated (FCM) fuel form are discussed. The matrix is produced via the nano-infiltration transient eutectic-forming (NITE) process. Coefficient of thermal expansion, thermal conductivity, and strength characteristics of this SiC matrix have been characterized in the unirradiated state. An ad hoc methodology for estimation of thermal conductivity of the neutron-irradiated NITE–SiC matrix is also provided to aid fuel performance modeling efforts specific to this concept. Finally, specific processing methods developed for production of an optimal and reliable fuel form using this process are summarized. These various sections collectively report the progress made to date on production of optimal FCM fuel form to enable its application in light water and advanced reactors

Advanced fuel for fast breeder reactors: Fabrication and properties and their optimization

International Nuclear Information System (INIS)

1988-06-01

The present design for FBR fuel rods includes usually MOX fuel pellets cladded into stainless steel tubes, together with UO 2 axial blanket and stainless steel hexagonal wrappers. Mixed carbide, nitride and metallic fuels have been tested as alternative fuels in test reactors. Among others, the objectives to develop these alternative fuels are to gain a high breeding ratio, short doubling time and high linear ratings. Fuel rod and assembly designers are now concentrating on finding the combination of optimized fuel, cladding and wrapper materials which could result in improvement of fuel operational reliability under high burnups and load-follow mode of operation. The purpose of the meeting was to review the experience of advanced FBR fuel fabrication technology, its properties before, under and after irradiation, peculiarities of the back-end of the nuclear fuel cycle, and to outline future trends. As a result of the panel discussion, the recommendations on future Agency activities in the area of advanced FBR fuels were developed. A separate abstract was prepared for each of the 10 presentations of this meeting. Refs, figs and tabs

AN ADVANCED OXIDATION PROCESS : FENTON PROCESS

Directory of Open Access Journals (Sweden)

Engin GÜRTEKİN

2008-03-01

Full Text Available Biological wastewater treatment is not effective treatment method if raw wastewater contains toxic and refractory organics. Advanced oxidation processes are applied before or after biological treatment for the detoxification and reclamation of this kind of wastewaters. The advanced oxidation processes are based on the formation of powerful hydroxyl radicals. Among advanced oxidation processes Fenton process is one of the most promising methods. Because application of Fenton process is simple and cost effective and also reaction occurs in a short time period. Fenton process is applied for many different proposes. In this study, Fenton process was evaluated as an advanced oxidation process in wastewater treatment.

Recent Advances in High-Performance Direct Methanol Fuel Cells

Science.gov (United States)

Narayanan, S. R.; Chun, W.; Valdez, T. I.; Jeffries-Nakamura, B.; Frank, H.; Surumpudi, S.; Halpert, G.; Kosek, J.; Cropley, C.; La Conti, A. B.;

Safety aspects of advanced fuels irradiations in EBR-II

International Nuclear Information System (INIS)

Lehto, W.K.

1975-09-01

Basic safety questions such as MFCI, loss-of-Na bond, pin behavior during design basis transients, and failure propagation were evaluated as they pertain to advanced fuels in EBR-II. With the exception of pin response to the unlikely loss-of-flow transient, the study indicates that irradiation of significant numbers of advanced fueled subassemblies in EBR-II should pose no safety problems. The analysis predicts, however, that Na boiling may occur during the postulated design basis unlikely loss-of-flow transient in subassemblies containing He-bonded fuel pins with the larger fuel-clad gaps. The calculations indicate that coolant temperatures at top of core in the limiting S/A's, containing the He bonded pins, would reach approximately 1480 0 F during the transient without application of uncertainty factors. Inclusion of uncertainties could result in temperature predictions which approach coolant boiling temperatures (1640 0 F). Further analysis of He-bonded pins is being done in this potential problem area, e.g., to apply best estimates of uncertainty factors and to determine the sensitivity of the preliminary results to gap conductance

Hydrogen-bromine fuel cell advance component development

Science.gov (United States)

Charleston, Joann; Reed, James

1988-01-01

Advanced cell component development is performed by NASA Lewis to achieve improved performance and longer life for the hydrogen-bromine fuel cells system. The state-of-the-art hydrogen-bromine system utilizes the solid polymer electrolyte (SPE) technology, similar to the SPE technology developed for the hydrogen-oxygen fuel cell system. These studies are directed at exploring the potential for this system by assessing and evaluating various types of materials for cell parts and electrode materials for Bromine-hydrogen bromine environment and fabricating experimental membrane/electrode-catalysts by chemical deposition.

Ultraclean Fuels Production and Utilization for the Twenty-First Century: Advances toward Sustainable Transportation Fuels

Energy Technology Data Exchange (ETDEWEB)

Fox, Elise B.; Liu, Zhong-Wen; Liu, Zhao-Tie

2013-11-21

Ultraclean fuels production has become increasingly important as a method to help decrease emissions and allow the introduction of alternative feed stocks for transportation fuels. Established methods, such as Fischer-Tropsch, have seen a resurgence of interest as natural gas prices drop and existing petroleum resources require more intensive clean-up and purification to meet stringent environmental standards. This review covers some of the advances in deep desulfurization, synthesis gas conversion into fuels and feed stocks that were presented at the 245th American Chemical Society Spring Annual Meeting in New Orleans, LA in the Division of Energy and Fuels symposium on "Ultraclean Fuels Production and Utilization".

EDF advanced fuel management strategies for the next century

International Nuclear Information System (INIS)

Kocher, A.; Charmensat, P.; Larderet, M.

1999-01-01

The French nuclear fleet represents 57 PWRs in operation, accounting for 80 % of France's total electricity production. The performance achieved by EDF reactors, in terms of availability (82.6% in 1997) and good cost control, have allowed to improve the nuclear KWh cost by 2% since 1992. The implementation of longer fuel cycles on the 1300 MW reactors from 1996 has contributed to this improvement and, as competitiveness is one of the main challenges for EDF, improving core management strategies is still at the order of the day. With this aim, a thinking process has been initiated to evaluate the benefit brought by the use of a fuel assembly like ALLIANCE, the new fuel product developed by Framatome-Fragema and FCF (Framatome Cogema Fuels) in close cooperation with EDF. The considered product provides enhanced performance, particularly as regards discharge burnup (at least up to 70 GWd/t) and thermal-hydraulic and mechanical behaviour. Fuel management improvements rely on the expertise gained by Framatome through designing core management strategies in a wide range of operating conditions prevailing in nuclear reactors all over the world. It will however be taken into account the necessity for EDF to adopt a policy of stepwise change owing to the potential impact of a 'series effect' on its numerous units. The proposed paper will describe innovative fuel managements, achievable thanks to advanced fuel assembly performance, that are jointly investigated by EDF and Framatome. It includes the following optimization schemes: extending cycle length by using higher enrichments up to 5%, while keeping the same reload size (1/3 core for example for the 1300 MW reactors); decreasing reload size (from 1/3 to 1/4 core), while keeping the same cycle length, using more enriched (up to 5 %) fuel assemblies; reaching annual cycle, with maximization of fuel cycle cost optimization (1/5 core). Beyond such schemes, combinations of optimized loading patterns and neutronic features of

Development of the advanced PHWR technology -Design and analysis of CANDU advanced fuel-

Energy Technology Data Exchange (ETDEWEB)

Suk, Hoh Chun; Shim, Kee Sub; Byun, Taek Sang; Park, Kwang Suk; Kang, Heui Yung; Kim, Bong Kee; Jung, Chang Joon; Lee, Yung Wook; Bae, Chang Joon; Kwon, Oh Sun; Oh, Duk Joo; Im, Hong Sik; Ohn, Myung Ryong; Lee, Kang Moon; Park, Joo Hwan; Lee, Eui Joon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

This is the `94 annual report of the CANDU advanced fuel design and analysis project, and describes CANFLEX fuel design and mechanical integrity analysis, reactor physics analysis and safety analysis of the CANDU-6 with the CANFLEX-NU. The following is the R and D scope of this fiscal year : (1) Detail design of CANFLEX-NU and detail analysis on the fuel integrity, reactor physics and safety. (a) Detail design and mechanical integrity analysis of the bundle (b) CANDU-6 refueling simulation, and analysis on the Xe transients and adjuster system capability (c) Licensing strategy establishment and safety analysis for the CANFLEX-NU demonstration demonstration irradiation in a commercial CANDU-6. (2) Production and revision of CANFLEX-NU fuel design documents (a) Production and approval of CANFLEX-NU reference drawing, and revisions of fuel design manual and technical specifications (b) Production of draft physics design manual. (3) Basic research on CANFLEX-SEU fuel. 55 figs, 21 tabs, 45 refs. (Author).

Computational Design of Advanced Nuclear Fuels

International Nuclear Information System (INIS)

Savrasov, Sergey; Kotliar, Gabriel; Haule, Kristjan

2014-01-01

The objective of the project was to develop a method for theoretical understanding of nuclear fuel materials whose physical and thermophysical properties can be predicted from first principles using a novel dynamical mean field method for electronic structure calculations. We concentrated our study on uranium, plutonium, their oxides, nitrides, carbides, as well as some rare earth materials whose 4f eletrons provide a simplified framework for understanding complex behavior of the f electrons. We addressed the issues connected to the electronic structure, lattice instabilities, phonon and magnon dynamics as well as thermal conductivity. This allowed us to evaluate characteristics of advanced nuclear fuel systems using computer based simulations and avoid costly experiments.

Advanced and sustainable fuel cycles for innovative reactor systems

International Nuclear Information System (INIS)

Glatz, J. P.; Malmbeck, R.; Purroy, D. S.; Soucek, P.; Inoue, T.; Uozumi, K.

2007-01-01

The key objective of nuclear energy systems of the future as defined by the Generation IV road map is to provide a sustainable energy generation for the future. It includes the requirement to minimize the nuclear waste produced and thereby notably reduce the long term stewardship burden in the future. It is therefore evident that the corresponding fuel cycles will play a central role in trying to achieve these goals by creating clean waste streams which contain almost exclusively the fission products. A new concept based on a grouped separation of actinides is widely discussed in this context, but it is of course a real challenge to achieve this type of separation since technologies available today have been developed to separate actinides from each other. In France, the CEA has launched extensive research programs in the ATALANTE facility in Marcoule to develop the advanced fuel cycles for new generation reactor systems. In this so called global actinide management (GAM) concept, the actinides are extracted in a sequence of chemical reactions (grouped actinide extraction (GANEX)) and immediately reintroduced in the fuel fabrication process is to use all actinides in the energy production process. The new group separation processes can be derived as in this case from aqueous techniques but also from so-called pyrochemical partitioning processes. Significant progress was made in recent years for both routes in the frame of the European research projects PARTNEW, PYROREP and EUROPART, mainly devoted to the separation of minor actinides in the frame of partitioning and transmutation (P and T) studies. The fuels used in the new generation reactors will be significantly different from the commercial fuels of today. Because of the fuel type and the very high burn-ups reached, pyrometallurgical reprocessing could be the preferred method. The limited solubility of some of the fuel materials in acidic aqueous solutions, the possibility to have an integrated irradiation and

Modeling constituent redistribution in U–Pu–Zr metallic fuel using the advanced fuel performance code BISON

International Nuclear Information System (INIS)

Galloway, J.; Unal, C.; Carlson, N.; Porter, D.; Hayes, S.

2015-01-01

Highlights: • An improved constituent distribution formulation in metallic nuclear fuels. • The new algorithm is implemented into the advanced fuel performance framework BISON. • Experimental Breeder Reactor-II data, T179, DP16, T459 are reanalyzed. • Phase dependent diffusion coefficients are improved. • Most influential phase is gamma, followed by alpha and thirdly the beta phase. - Abstract: An improved robust formulation for constituent distribution in metallic nuclear fuels is developed and implemented into the advanced fuel performance framework BISON. The coupled thermal diffusion equations are solved simultaneously to reanalyze the constituent redistribution in post irradiation data from fuel tests performed in Experimental Breeder Reactor-II (EBR-II). Deficiencies observed in previously published formulation and numerical implementations are also improved. The present model corrects an inconsistency between the enthalpies of solution and the solubility limit curves of the phase diagram while also adding an artificial diffusion term when in the 2-phase regime that stabilizes the standard Galerkin finite element (FE) method used by BISON. An additional improvement is in the formulation of zirconium flux as it relates to the Soret term. With these new modifications, phase dependent diffusion coefficients are revaluated and compared with the previously recommended values. The model validation included testing against experimental data from fuel pins T179, DP16 and T459, irradiated in EBR-II. A series of viable material properties for U–Pu–Zr based materials was determined through a sensitivity study, which resulted in three cases with differing parameters that showed strong agreement with one set of experimental data, rod T179. Subsequently a full-scale simulation of T179 was performed to reduce uncertainties, particularly relating to the temperature boundary condition for the fuel. In addition a new thermal conductivity model combining all

Advances in HTGR fuel performance models

International Nuclear Information System (INIS)

Stansfield, O.M.; Goodin, D.T.; Hanson, D.L.; Turner, R.F.

1985-01-01

Advances in HTGR fuel performance models have improved the agreement between observed and predicted performance and contributed to an enhanced position of the HTGR with regard to investment risk and passive safety. Heavy metal contamination is the source of about 55% of the circulating activity in the HTGR during normal operation, and the remainder comes primarily from particles which failed because of defective or missing buffer coatings. These failed particles make up about 5 x 10 -4 fraction of the total core inventory. In addition to prediction of fuel performance during normal operation, the models are used to determine fuel failure and fission product release during core heat-up accident conditions. The mechanistic nature of the models, which incorporate all important failure modes, permits the prediction of performance from the relatively modest accident temperatures of a passively safe HTGR to the much more severe accident conditions of the larger 2240-MW/t HTGR. (author)

Ternary carbide uranium fuels for advanced reactor design applications

International Nuclear Information System (INIS)

Knight, Travis; Anghaie, Samim

1999-01-01

Solid-solution mixed uranium/refractory metal carbides such as the pseudo-ternary carbide, (U, Zr, Nb)C, hold significant promise for advanced reactor design applications because of their high thermal conductivity and high melting point (typically greater than 3200 K). Additionally, because of their thermochemical stability in a hot-hydrogen environment, pseudo-ternary carbides have been investigated for potential space nuclear power and propulsion applications. However, their stability with regard to sodium and improved resistance to attack by water over uranium carbide portends their usefulness as a fuel for advanced terrestrial reactors. An investigation into processing techniques was conducted in order to produce a series of (U, Zr, Nb)C samples for characterization and testing. Samples with densities ranging from 91% to 95% of theoretical density were produced by cold pressing and sintering the mixed constituent carbides at temperatures as high as 2650 K. (author)

RU fuel development program for an advanced fuel cycle in Korea

International Nuclear Information System (INIS)

Suk, Hochum; Sim, Kiseob; Kim, Bongghi; Inch, W.W.; Page, R.

1998-01-01

Korea is a unique country, having both PWR and CANDU reactors. Korea can therefore exploit the natural synergism between the two reactor types to minimize overall waste production, and maximize energy derived from the fuel, by ultimately burning the spent fuel from its PWR reactors in CANDU reactors. As one of the possible fuel cycles, Recovered Uranium (RU) fuel offers a very attractive alternative to the use of Natural Uranium (NU) and slightly enriched uranium (SEU) in CANDU reactors. Potential benefits can be derived from a number of stages in the fuel cycle: no enrichment required, therefore no enrichment tails, direct conversion to UO 2 , lower sensitivity to 234 U and 236U absorption in the CANDU reactor, and expected lower cost relative to NU and SEU. These benefits all fit well with the PWR-CANDU fuel cycle synergy. RU arising from the conventional reprocessing of European and Japanese oxide spent fuel by 2000 is projected to be approaching 25,000 te. The use of RU fuel in a CANDU 6 reactor should result in no serious radiological difficulties and no requirements for special precautions and should not require any new technologies for the fuel fabrication and handling. The use of the CANDU Flexible Fueling (CANFLEX) bundle as the carrier for RU will be fully compatible with the reactor design, current safety and operational requirements, and there will be improved fuel performance compared with the CANDU 37-element NU fuel bundle. Compared with the 37-element NU bundle, the RU fuel has significantly improved fuel cycle economics derived from increased burnups, a large reduction in both fuel requirements and spent fuel, arisings, and the potential lower cost for RU material. There is the potential for annual fuel cost savings in the range of one-third to two-thirds, with enhanced operating margins using RU in the CANFLEX bundle design. These benefits provide the rationale for justifying R and D efforts on the use of RU fuel for advanced fuel cycles in CANDU

VVANTAGE 6 - an advanced fuel assembly design for VVER reactors

International Nuclear Information System (INIS)

Doshi, P.K.; DeMario, E.E.; Knott, R.P.

1993-01-01

Over the last 25 years, Westinghouse fuel assemblies for pressurized water reactors (PWR's) have undergone significant changes to the current VANTAGE 5. VANTAGE 5 PWR fuel includes features such as removable top nozzles, debris filter bottom nozzles, low-pressure-drop zircaloy grids, zircaloy intermediate flow mixing grids, optimized fuel rods, in-fuel burnable absorbers, and increased burnup capability to region average values of 48000 MWD/MTU. These features have now been adopted to the VVER reactors. Westinghouse has completed conceptual designs for an advanced fuel assembly and other core components for VVER-1000 reactors known as VANTAGE 6. This report describes the VVANTAGE 6 fuel assembly design

Alternative Fuel and Advanced Technology Commercial Lawn Equipment

Energy Technology Data Exchange (ETDEWEB)

None

2014-10-10

The U.S. Department of Energy's Clean Cities program produced this guide to help inform the commercial mowing industry about product options and potential benefits. This guide provides information about equipment powered by propane, ethanol, compressed natural gas, biodiesel, and electricity, as well as advanced engine technology. In addition to providing an overview for organizations considering alternative fuel lawn equipment, this guide may also be helpful for organizations that want to consider using additional alternative fueled equipment.

Advancing liquid metal reactor technology with nitride fuels

International Nuclear Information System (INIS)

Lyon, W.F.; Baker, R.B.; Leggett, R.D.; Matthews, R.B.

1991-08-01

A review of the use of nitride fuels in liquid metal fast reactors is presented. Past studies indicate that both uranium nitride and uranium/plutonium nitride possess characteristics that may offer enhanced performance, particularly in the area of passive safety. To further quantify these effects, the analysis of a mixed-nitride fuel system utilizing the geometry and power level of the US Advanced Liquid Metal Reactor as a reference is described. 18 refs., 2 figs., 2 tabs

Research on CDA for advanced fuel FBR

International Nuclear Information System (INIS)

Hirano, Go; Hirakawa, Naohiro; Kawada, Ken-ichi; Niwa, Hazime.

1997-03-01

For the purpose of evaluating possibility of the re-criticality of a metallic fueled reactor, Tohoku university and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled 'Research on CDA for advanced fuel FBR'. The results of this year are the following. The accident initiator considered is a loss-of-flow accident with ATWS. The LOF analysis was performed for the metallic fueled 600 MWe homogeneous two region reactors, both for a metallic fuel only and for a metallic fuel core with ZrH pin. The SAS3D CDA initiation phase analysis code was used to investigate the re-criticality potential at the severe accident. The change mainly in the constants was necessary to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. LOF with flow decay half time of t 1/2 =0.5(s) (all blackout case) and 5.5(s) (ordinary LOF case) were analyzed. Independent of the conditions of the analysis, the results show all the cases could avoid to become prompt-critical. Depending on the analysis condition, it becomes necessary to transfer to the transient phase, it is also shown there is a possibility to avoid re-criticality due to the motion of molten fuel both for the metallic fuel and for the metallic fuel with ZrH moderator. However, because of the constants used for the material property the results might overestimate the fuel motion. It is shown that the moderator is effective to terminate the accident at an early stage. The behavior of metallic fueled reactors at CDA was analyzed with SAS3D code by modifying the constants of material properties to be applied to the reactor. It is shown that a metallic fueled reactor has a possibility to avoid re-criticality at CDA. (J.P.N.)

Results of trial operation of the WWER advanced fuel assemblies

International Nuclear Information System (INIS)

Vasilchenko, I.; Dragunov, Y.; Mikhalchuk, A.

2001-01-01

The paper describes results from experimental operation of advanced WWER-1000 fuel assemblies (AFA) at five units in Balakovo NPP. Advanced fuel is developed according to the concept of standard WWER-1000 fuel assembly (jacket-free). The new features includes: 1) zirconium guiding channels (alloy E-635 and E-110) and spacer grids (alloy E-110); 2) integrated burnable absorber gadolinium; 3) extended service life of fuel assemblies (FA) and absorber rods (possibility of repair of FA); 4) improved adoption to reactor conditions. Some results of AFA pilot operation of a three year operation are presented and analyses of effectiveness of improvements are made concerning application of zirconium channels and grids; application of integrated burnable absorbers; extension of FA and absorbing rods service life and FA repairability. These new features of WWER-1000 fuel design allow: 1) to reduce the average fuel enrichment to the 3.77% instead of 4.31% in U-235; 2) to reduce the FA axial load in reactor hot state by 40%,; 3) increasing of fuel operation in reactor to the 30000 effective days with possibility to have a 5-year residence time in the reactor. The design of new generation FA for WWER-440 reactors involves few key changes. Fuel inventory in new fuel design is increased due to elongation of fuel stack and reducing the diameter of the central hole. Vibration stability is enhanced as a result of: no-play junction of the fuel rod with the lower grid; change of SG arrangements; strengthening of the lower grid unit; secure of the central tube in the gap. Water-uranium ration is increased. Introduction of all these kinds of modernization in a 5-year fuel cycle reduces fuel component in the energy cost to the 7%

Advances in fuel cell vehicle design

Science.gov (United States)

Bauman, Jennifer

Factors such as global warming, dwindling fossil fuel reserves, and energy security concerns combine to indicate that a replacement for the internal combustion engine (ICE) vehicle is needed. Fuel cell vehicles have the potential to address the problems surrounding the ICE vehicle without imposing any significant restrictions on vehicle performance, driving range, or refuelling time. Though there are currently some obstacles to overcome before attaining the widespread commercialization of fuel cell vehicles, such as improvements in fuel cell and battery durability, development of a hydrogen infrastructure, and reduction of high costs, the fundamental concept of the fuel cell vehicle is strong: it is efficient, emits zero harmful emissions, and the hydrogen fuel can be produced from various renewable sources. Therefore, research on fuel cell vehicle design is imperative in order to improve vehicle performance and durability, increase efficiency, and reduce costs. This thesis makes a number of key contributions to the advancement of fuel cell vehicle design within two main research areas: powertrain design and DC/DC converters. With regards to powertrain design, this research first analyzes various powertrain topologies and energy storage system types. Then, a novel fuel cell-battery-ultracapacitor topology is presented which shows reduced mass and cost, and increased efficiency, over other promising topologies found in the literature. A detailed vehicle simulator is created in MATLAB/Simulink in order to simulate and compare the novel topology with other fuel cell vehicle powertrain options. A parametric study is performed to optimize each powertrain and general conclusions for optimal topologies, as well as component types and sizes, for fuel cell vehicles are presented. Next, an analytical method to optimize the novel battery-ultracapacitor energy storage system based on maximizing efficiency, and minimizing cost and mass, is developed. This method can be applied

Influence of prolonged nuclear fuel burnup on safety characteristics of advanced PWRs

International Nuclear Information System (INIS)

Spasojevic, D.; Matausek, M.; Marinkovic, N.

1989-01-01

Prolonged nuclear fuel burnup in advanced NPP with four or more instead of three one-year cycles, and/or with 15- to 18-month instead of standard 12-month cycles, requires the fresh fuel to have increased enrichment combined with burnable poisons. This causes changes in axial and radial distribution of power generation during the particular fuel cycles, so that detailed analyses of thermal reliability of reactor core becomes necessary. This paper presents the results of the analysis of the departure from nuclear boiling ratio DNBR for an equilibrium cycle of an advanced PWR. (author)

Advanced Fuel Pellet Materials and Fuel Rod Design for Water Cooled Reactors. Proceedings of a Technical Committee Meeting

International Nuclear Information System (INIS)

2010-10-01

The economics of current nuclear power plants have improved through increased fuel burnup and longer fuel cycles, i.e. increasing the effective time that fuel remains in the reactor core and the amount of energy it generates. Efficient consumption of fissile material in the fuel element before it is discharged from the reactor means that less fuel is required over the reactor's life cycle, which results in lower amounts of fresh fuel, lower spent fuel storage costs, and less waste for ultimate disposal. Better utilization of fissile nuclear materials, as well as more flexible power manoeuvring, place challenging operational demands on materials used in reactor components, and first of all, on fuel and cladding materials. It entails increased attention to measures ensuring desired in-pile fuel performance parameters that require adequate improvements in fuel material properties and fuel rod designs. These are the main reasons that motivated the IAEA Technical Working Group on Fuel Performance and Technology (TWG-FPT) to recommend the organization of a Technical Committee Meeting on Advanced Fuel Pellet Materials and Fuel Rod Designs for Power Reactors. The proposal was supported by the IAEA TWGs on Advanced Technologies for Light and Heavy Water-Cooled Reactors (TWG-LWR and TWG-HWR), and the meeting was held at the invitation of the Government of Switzerland at the Paul Scherrer Institute in Villigen, from 23 to 26 November 2009. This was the third IAEA meeting on these subjects (the first was held in 1996 in Tokyo, Japan, and the second in 2003 in Brussels, Belgium), which reflects the continuous interest in the above issues among Member States. The purpose of the meeting was to review the current status in the development of fuel pellet materials and to explore recent improvements in fuel rod designs for light and heavy water cooled power reactors. The meeting was attended by 45 specialists representing fuel vendors, nuclear utilities, research and development

The advanced fuel cycle facility (AFCF) role in the global nuclear energy partnership

International Nuclear Information System (INIS)

Griffith, Andrew

2007-01-01

The Global Nuclear Energy Partnership (GNEP), launched in February, 2006, proposes to introduce used nuclear fuel recycling in the United States with improved proliferation-resistance and a more effective waste management approach. This program is evaluating ways to close the fuel cycle in a manner that builds on recent laboratory breakthroughs in U.S. national laboratories and draws on international and industry partnerships. Central to moving this advanced fuel recycling technology from the laboratory to commercial implementation is a flexible research, development and demonstration facility, called the Advanced Fuel Cycle Facility (AFCF). The AFCF was introduced as one of three projects under GNEP and will provide the U.S. with the capabilities to evaluate technologies that separate used fuel into reusable material and waste in a proliferation-resistant manner. The separations technology demonstration capability is coupled with a remote transmutation fuel fabrication demonstration capability in an integrated manner that demonstrates advanced safeguard technologies. This paper will discuss the key features of AFCF and its support of the GNEP objectives. (author)

Advances in AGR fuel fabrication - now and the future

International Nuclear Information System (INIS)

Bleasdale, P.A.

1995-01-01

To date, over 3 million AGR fuel pins have been manufactured at Springfields for the UK AGR programme. During this time, AGR fuel design and manufacture has developed and evolved in response to the needs of the reactor operators to enhance fuel reliability and performance. More recently, major advances have been made in the systems and organisational culture which support fuel manufacture at Fuel Division. The introduction of MRP II in 1989 into Fuel Division enabled significant reductions in stock and work-in-progress, together with reductions in manufacturing lead times. Other successful initiatives introduced into Fuel Division have been Just-in-Time (JIT) and AST (Additional Skills Training) which have built on the success of MRP II. All of these initiatives are evidence of Fuel Division's ''Total Quality'' approach to fabricating fuel. Fuel Division is currently in the final stages of commissioning the New Oxide Fuels Complex (NOFC) where both AGR and PWR fuel will be manufactured to the highest standards of quality, safety and environmental protection. NOFC is a totally integrated plant which represents a Pound 200M investment, demonstrating Fuel Division's commitment to building on its 40+ years of fuel fabrication experience and ensuring secure supply of fuel to its customers for years to come. (author)

Advance reactor and fuel-cycle systems--potentials and limitations for United States utilities

International Nuclear Information System (INIS)

Zebroski, E.L.; Williams, R.F.

1979-01-01

This paper reviews the potential benefits and limitations of advance reactor and fuel-cycle systems for United States utilities. The results of the review of advanced technologies show that for the near and midterm, the only advance reactor and fuel-cycle system with significant potential for United States utilities is the current LWR, and evolutionary, not revolutionary, enhancements. For the long term, the liquid-metal breeder reactor continues to be the most promising advance nuclear option. The major factors leading to this conclusion are summarized

Modified-open fuel cycle performance with breed-and-burn advanced reactor concepts

International Nuclear Information System (INIS)

Heidet, Florent; Kim, Taek K.; Taiwo, Temitope A.

2011-01-01

Recent advances in fast reactor designs enable significant increase in the uranium utilization in an advanced fuel cycle. The category of fast reactors, collectively termed breed-and-burn reactor concepts, can use a large amount of depleted uranium as fuel without requiring enrichment with the exception of the initial core critical loading. Among those advanced concepts, some are foreseen to operate within a once-through fuel cycle such as the Traveling Wave Reactor, CANDLE reactor or Ultra-Long Life Fast Reactor, while others are intended to operate within a modified-open fuel cycle, such as the Breed-and-Burn reactor and the Energy Multiplier Module. This study assesses and compares the performance of the latter category of breed-and-burn reactors at equilibrium state. It is found that the two reactor concepts operating within a modified-open fuel cycle can significantly improve the sustainability and security of the nuclear fuel cycle by decreasing the uranium resources and enrichment requirements even further than the breed-and-burn core concepts operating within the once-through fuel cycle. Their waste characteristics per unit of energy are also found to be favorable, compared to that of currently operating PWRs. However, a number of feasibility issues need to be addressed in order to enable deployment of these breed-and-burn reactor concepts. (author)

Advanced combinational microfluidic multiplexer for fuel cell reactors

International Nuclear Information System (INIS)

Lee, D W; Kim, Y; Cho, Y-H; Doh, I

2013-01-01

An advanced combinational microfluidic multiplexer capable to address multiple fluidic channels for fuel cell reactors is proposed. Using only 4 control lines and two different levels of control pressures, the proposed multiplexer addresses up to 19 fluidic channels, at least two times larger than the previous microfluidic multiplexers. The present multiplexer providing high control efficiency and simple structure for channel addressing would be used in the application areas of the integrated microfluidic systems such as fuel cell reactors and dynamic pressure generators

Cost reductions of fuel cells for transport applications: fuel processing options

Energy Technology Data Exchange (ETDEWEB)

Teagan, W P; Bentley, J; Barnett, B [Arthur D. Little, Inc., Cambridge, MA (United States)

1998-03-15

The highly favorable efficiency/environmental characteristics of fuel cell technologies have now been verified by virtue of recent and ongoing field experience. The key issue regarding the timing and extent of fuel cell commercialization is the ability to reduce costs to acceptable levels in both stationary and transport applications. It is increasingly recognized that the fuel processing subsystem can have a major impact on overall system costs, particularly as ongoing R and D efforts result in reduction of the basic cost structure of stacks which currently dominate system costs. The fuel processing subsystem for polymer electrolyte membrane fuel cell (PEMFC) technology, which is the focus of transport applications, includes the reformer, shift reactors, and means for CO reduction. In addition to low cost, transport applications require a fuel processor that is compact and can start rapidly. This paper describes the impact of factors such as fuel choice operating temperature, material selection, catalyst requirements, and controls on the cost of fuel processing systems. There are fuel processor technology paths which manufacturing cost analyses indicate are consistent with fuel processor subsystem costs of under $150/kW in stationary applications and $30/kW in transport applications. As such, the costs of mature fuel processing subsystem technologies should be consistent with their use in commercially viable fuel cell systems in both application categories. (orig.)

Alternative Fuels and Advanced Vehicles: Resources for Fleet Managers (Clean Cities) (Presentation)

Energy Technology Data Exchange (ETDEWEB)

Brennan, A.

2011-04-01

A discussion of the tools and resources on the Clean Cities, Alternative Fuels and Advanced Vehicles Data Center, and the FuelEconomy.gov Web sites that can help vehicle fleet managers make informed decisions about implementing strategies to reduce gasoline and diesel fuel use.

Advanced diagnostics in oxy-fuel combustion processes

Energy Technology Data Exchange (ETDEWEB)

Brix, J.; Clausen, Soennik; Degn Jensen, A. (Technical Univ. of Denmark. CHEC Research Centre, Kgs. Lyngby (Denmark)); Boeg Toftegaard, M. (DONG Energy Power, Hvidovre (Denmark))

2012-07-01

This report sums up the findings in PSO-project 010069, ''Advanced Diagnostics in Oxy-Fuel Combustion Processes''. Three areas of optic diagnostics are covered in this work: - FTIR measurements in a 30 kW swirl burner. - IR measurements in a 30 kW swirl burner. - IR measurements in a laboratory scale fixed bed reactor. The results obtained in the swirl burner have proved the FTIR method as a valuable technique for gas phase temperature measurements. When its efficacy is evaluated against traditional thermocouple measurements, two cases, with and without probe beam stop, must however be treated separately. When the FTIR probe is operated with the purpose of gas phase concentration measurements the probe needs to operate with a beam stop mounted in front of it. With this beam stop in place it was shown that the measured gas phase temperature was affected by cooling, induced by the cooled beam stop. Hence, for a more accurate determination of gas phase temperatures the probe needed to operate without the beam stop. When this was the case, the FTIR probe showed superior to traditional temperature measurements using a thermocouple as it could measure the fast temperature fluctuations. With the beam stop in place the efficacy of the FTIR probe for gas temperature determination was comparable to the use of a traditional thermocouple. The evaluation of the FTIR technique regarding estimation of gas phase concentrations of H{sub 2}O, CO{sub 2} and CO showed that the method is reliable though it cannot be stated as particularly accurate. The accuracy of the method is dependent on the similarity of the reference emission spectra of the gases with those obtained in the experiments, as the transmittance intensity is not a linear function of concentration. The length of the optical path also affects the steadiness of the measurements. The length of the optical path is difficult to adjust on the small scales that are the focus of this work. However

Current status on advanced aqueous reprocessing process (next) in FaCT project

International Nuclear Information System (INIS)

Washiya, Tadahiro; Myochin, Munetaka; Koyama, Tomozo

2009-01-01

Japan Atomic Energy Agency (JAEA) launched the Fast Reactor Cycle Technology Development (FaCT) project in cooperation with the Japanese electric utilities in 2006. An integration of the advanced aqueous reprocessing concept and the simplified pelletizing fuel fabrication was selected as the most promising fuel cycle system. In order to accomplish the integration, R and D tasks were launched as FaCT Project in 2006 by Japanese joint team. The New Extraction System for TRU Recovery (NEXT) system is an advanced aqueous reprocessing concept which was based on the well established aqueous reprocessing for LWR spent fuel and newly applied processes such as uranium crystallization and extraction chromatography for MAs recovery. Main task of the NEXT process is to develop the TRU recovery process and equipments with high reliability, criticality safety, high durability and remote maintainability. In the FaCT project, all innovative technologies are planned to be developed within the next decade focusing on the future commercialization of FBR cycle systems. The judgment of the adoption of each innovative technology will be made by 2010 based on the results of R and Ds. The development of each technology is to be completed by around 2015. By the same time, it is scheduled to present the conceptual design of commercial and demonstrative fast reactor cycle facilities. The six items (Disassembling and shearing, Fuel dissolution, Uranium Crystallization, Single cycle co-extraction of U, Pu and Np, MA recovery by extraction chromatography and Waste treatment) have been identified as the issues to be developed corresponding to each process step. Current R and D status and prospects of this system until around 2015 is reported. (author)

Review and comment on the advanced spent fuel management process (1): Technical aspects and non-proliferation concerns

International Nuclear Information System (INIS)

Song, Yo Taik

2001-01-01

Efforts are made to analyze the project, the Advanced Spent Fuel Management Technology (ASFMT), which is currently carried out at Korea Atomic Energy Research Institute, on the technical feasibility and validity as well as on the nuclear non-proliferation concerns. The project is a part of a program under the 'Long and Midterm Nuclear Development Program'. On the technical analysis, reviewed the papers presented at the national and international meetings on the subject by KAERI staffs, and also participated to various technical discussions on the 'Mock-up Test', currently in progress. On the non-proliferation concerns, the ASFMT project was reviewed and analyzed in reference to various programs currently in progress or in a formulation stages in US, such as the DOE TOPS and ATW. Further reviewed the past JASNEC process and programs for possible application of the ASFMT project for JASNEC project. Provided a few thoughts for effectively carrying out the ASFMT project, and a plan for the next phase is presented

Review and comment on the advanced spent fuel management process (1): Technical aspects and non-proliferation concerns

Energy Technology Data Exchange (ETDEWEB)

Song, Yo Taik

2001-01-01

Efforts are made to analyze the project, the Advanced Spent Fuel Management Technology (ASFMT), which is currently carried out at Korea Atomic Energy Research Institute, on the technical feasibility and validity as well as on the nuclear non-proliferation concerns. The project is a part of a program under the 'Long and Midterm Nuclear Development Program'. On the technical analysis, reviewed the papers presented at the national and international meetings on the subject by KAERI staffs, and also participated to various technical discussions on the 'Mock-up Test', currently in progress. On the non-proliferation concerns, the ASFMT project was reviewed and analyzed in reference to various programs currently in progress or in a formulation stages in US, such as the DOE TOPS and ATW. Further reviewed the past JASNEC process and programs for possible application of the ASFMT project for JASNEC project. Provided a few thoughts for effectively carrying out the ASFMT project, and a plan for the next phase is presented.

Effect of channel aspect ratio on chemical recuperation process in advanced aeroengines

International Nuclear Information System (INIS)

Zhang, Silong; Cui, Naigang; Xiong, Yuefei; Feng, Yu; Qin, Jiang; Bao, Wen

2017-01-01

The working process of an advanced aeroengine such as scramjet with endothermic hydrocarbon fuel cooling is a chemical recuperative cycle. The design of cooling channel in terms of engine real working conditions is very important for the chemical recuperation process. To study the effects of channel aspect ratio (AR) on chemical recuperation process of advanced aeroengines, three dimensional model of pyrolysis coolant flow inside asymmetrical rectangular cooling channels with fins is introduced and validated through experiments. Cases when AR varies from 1 to 8 are carried out. In the pyrolysis zone of the cooling channel, decreasing the channel aspect ratio can reduce the temperature difference and non-uniformity of fuel conversion in the channel cross section, and it can also increase the final conversion and corresponding chemical heat absorption. A small channel aspect ratio is beneficial for the chemical recuperation process and can guarantee the engine cooling performance in the pyrolysis zone of the cooling channel. - Highlights: • Large non-uniformity of conversion is bad for the chemical recuperation. • Small channel aspect ratio is beneficial for improving the chemical recuperation effectiveness. • Small channel aspect ratio is also beneficial for reducing the engine wall temperature.

Use of Friction Stir Welding and Friction Stir Processing for Advanced Nuclear Fuels and Materials Joining Applications

International Nuclear Information System (INIS)

J. I. Cole; J. F. Jue

2006-01-01

Application of the latest developments in materials technology may greatly aid in the successful pursuit of next generation reactor and transmutation technologies. One such area where significant progress is needed is joining of advanced fuels and materials. Rotary friction welding, also referred to as friction stir welding (FSW), has shown great promise as a method for joining traditionally difficult to join materials such as aluminum alloys. This relatively new technology, first developed in 1991, has more recently been applied to higher melting temperature alloys such as steels, nickel-based and titanium alloys. An overview of the FSW technology is provided and two specific nuclear fuels and materials applications where the technique may be used to overcome limitations of conventional joining technologies are highlighted

Study of advanced fuel system concepts for commercial aircraft and engines

Science.gov (United States)

Versaw, E. F.; Brewer, G. D.; Byers, W. D.; Fogg, H. W.; Hanks, D. E.; Chirivella, J.

1983-01-01

The impact on a commercial transport aircraft of using fuels which have relaxed property limits relative to current commercial jet fuel was assessed. The methodology of the study is outlined, fuel properties are discussed, and the effect of the relaxation of fuel properties analyzed. Advanced fuel system component designs that permit the satisfactory use of fuel with the candidate relaxed properties in the subject aircraft are described. The two fuel properties considered in detail are freezing point and thermal stability. Three candidate fuel system concepts were selected and evaluated in terms of performance, cost, weight, safety, and maintainability. A fuel system that incorporates insulation and electrical heating elements on fuel tank lower surfaces was found to be most cost effective for the long term.

Advanced immobilization processes for fuel hulls and dissolver residues

International Nuclear Information System (INIS)

Hebel, W.; Boehme, G.; Findlay, J.R.; Sombret, C.

1984-08-01

Various research and development projects for the conditioning of cladding scraps and dissolver residues are pursued within the scope of the R and D programme on nuclear waste Management of the European Community. They include the characterization of the waste materials arising from industrial fuel reprocessing and the development of different waste immobilization techniques. These concern the embedment of scraps and residues into inert matrices like cement, metal alloys, compacted graphite and sintered ceramics as well as the treatment of the fuel hulls by melting or chemical conversion. The conditioned waste forms are tested as to their relevant properties for activity enclosure

Advanced immobilization processes for fuel hulls and dissolver residues

International Nuclear Information System (INIS)

Hebel, W.; Boehme, G.; Findlay, J.R.; Sombert, C.

1984-01-01

Various research and development projects for the conditioning of cladding scraps and dissolver residues are pursued within the scope of the R and D programme on nuclear waste Management of the European Community. They include the characterization of the waste materials arising from industrial fuel reprocessing and the development of different waste immobilization techniques. These concern the embedment of the scraps and residues into inert matrices like cement, metal alloys, compacted graphite and sintered ceramics as well as the treatment of the fuel hulls by melting or chemical conversion. The conditioned waste forms are tested as to their relevant properties for activity enclosure

Advanced fuels for nuclear fusion reactors

International Nuclear Information System (INIS)

McNally, J.R. Jr.

1974-01-01

Should magnetic confinement of hot plasma prove satisfactory at high β (16 πnkT//sub B 2 / greater than 0.1), thermonuclear fusion fuels other than D.T may be contemplated for future fusion reactors. The prospect of the advanced fusion fuels D.D and 6 Li.D for fusion reactors is quite promising provided the system is large, well reflected and possesses a high β. The first generation reactions produce the very active, energy-rich fuels t and 3 He which exhibit a high burnup probability in very hot plasmas. Steady state burning of D.D can ensue in a 60 kG field, 5 m reactor for β approximately 0.2 and reflectivity R/sub mu/ = 0.9 provided the confinement time is about 38 sec. The feasibility of steady state burning of 6 Li.D has not yet been demonstrated but many important features of such systems still need to be incorporated in the reactivity code. In particular, there is a need for new and improved nuclear cross section data for over 80 reaction possibilities

Nuclear Fuel Cycle Analysis Technology to Develop Advanced Nuclear Fuel Cycle

Energy Technology Data Exchange (ETDEWEB)

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

2011-12-15

The nuclear fuel cycle (NFC) analysis is a study to set a NFC policy and to promote systematic researches by analyzing technologies and deriving requirements at each stage of a fuel cycle. System analysis techniques are utilized for comparative analysis and assessment of options on a considered system. In case that NFC is taken into consideration various methods of the system analysis techniques could be applied depending on the range of an interest. This study presented NFC analysis strategies for the development of a domestic advanced NFC and analysis techniques applicable to different phases of the analysis. Strategically, NFC analysis necessitates the linkage with technology analyses, domestic and international interests, and a national energy program. In this respect, a trade-off study is readily applicable since it includes various aspects on NFC as metrics and then analyzes the considered NFC options according to the derived metrics. In this study, the trade-off study was identified as a method for NFC analysis with the derived strategies and it was expected to be used for development of an advanced NFC. A technology readiness level (TRL) method and NFC simulation codes could be utilized to obtain the required metrics and data for assessment in the trade-off study. The methodologies would guide a direction of technology development by comparing and assessing technological, economical, environmental, and other aspects on the alternatives. Consequently, they would contribute for systematic development and deployment of an appropriate advanced NFC.

Nuclear Fuel Cycle Analysis Technology to Develop Advanced Nuclear Fuel Cycle

International Nuclear Information System (INIS)

Park, Byung Heung; Ko, Won IL

2011-01-01

The nuclear fuel cycle (NFC) analysis is a study to set a NFC policy and to promote systematic researches by analyzing technologies and deriving requirements at each stage of a fuel cycle. System analysis techniques are utilized for comparative analysis and assessment of options on a considered system. In case that NFC is taken into consideration various methods of the system analysis techniques could be applied depending on the range of an interest. This study presented NFC analysis strategies for the development of a domestic advanced NFC and analysis techniques applicable to different phases of the analysis. Strategically, NFC analysis necessitates the linkage with technology analyses, domestic and international interests, and a national energy program. In this respect, a trade-off study is readily applicable since it includes various aspects on NFC as metrics and then analyzes the considered NFC options according to the derived metrics. In this study, the trade-off study was identified as a method for NFC analysis with the derived strategies and it was expected to be used for development of an advanced NFC. A technology readiness level (TRL) method and NFC simulation codes could be utilized to obtain the required metrics and data for assessment in the trade-off study. The methodologies would guide a direction of technology development by comparing and assessing technological, economical, environmental, and other aspects on the alternatives. Consequently, they would contribute for systematic development and deployment of an appropriate advanced NFC.

Issues of high-burnup fuel for advanced nuclear reactors

International Nuclear Information System (INIS)

Belac, J.; Milisdoerfer, L.

2004-12-01

A brief description is given of nuclear fuels for Generation III+ and IV reactors, and the major steps needed for a successful implementation of new fuels in prospective types of newly designed power reactors are outlined. The following reactor types are discussed: gas cooled fast reactors, heavy metal (lead) cooled fast reactors, molten salt cooled reactors, sodium cooled fast reactors, supercritical water cooled reactors, and very high temperature reactors. The following are regarded as priority areas for future investigations: (i) spent fuel radiotoxicity; (ii) proliferation volatility; (iii) neutron physics characteristics and inherent safety element assessment; technical and economic analysis of the manufacture of advanced fuels; technical and economic analysis of the fuel cycle back end, possibilities of spent nuclear fuel reprocessing, storage and disposal. In parallel, work should be done on the validation and verification of analytical tools using existing and/or newly acquired experimental data. (P.A.)

A review on the development of the advanced fuel fabrication technology

International Nuclear Information System (INIS)

Lee, Jung Won; Lee, Yung Woo; Sohn, Dong Sung; Yang, Myung Seung; Bae, Kee Kwang; Nah, Sang Hoh; Kim, Han Soo; Kim, Bong Koo; Song, Keun Woo; Kim, See Hyung

1995-07-01

In this state-of art report, the development status of the advanced nuclear fuel was investigated. The current fabrication technology for coated particle fuel and non-oxide fuel such as sol-gel technology, coating technology, and carbothermic reduction reaction has also been examined. In the view point of inherent safety and efficiency in the operation of power plant, the coated particle fuel will keep going on its reputation as nuclear fuel for a high temperature gas cooled reactor, and the nitride fuel is very prospective for the next liquid metal fast breeder reactor. 43 figs., 17 tabs., 96 refs. (Author)

A review on the development of the advanced fuel fabrication technology

Energy Technology Data Exchange (ETDEWEB)

Lee, Jung Won; Lee, Yung Woo; Sohn, Dong Sung; Yang, Myung Seung; Bae, Kee Kwang; Nah, Sang Hoh; Kim, Han Soo; Kim, Bong Koo; Song, Keun Woo; Kim, See Hyung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

In this state-of art report, the development status of the advanced nuclear fuel was investigated. The current fabrication technology for coated particle fuel and non-oxide fuel such as sol-gel technology, coating technology, and carbothermic reduction reaction has also been examined. In the view point of inherent safety and efficiency in the operation of power plant, the coated particle fuel will keep going on its reputation as nuclear fuel for a high temperature gas cooled reactor, and the nitride fuel is very prospective for the next liquid metal fast breeder reactor. 43 figs., 17 tabs., 96 refs. (Author).

Advanced Nuclear Fuels Corporation: one year later

International Nuclear Information System (INIS)

Bjoernard, T.A.; Sofer, G.A.

1988-01-01

About one year ago, after 18 years of business as a wholly owned affiliate of Exxon Corporation, Exxon Nuclear Company was acquired by Siemens/KWU and its name was changed to Advanced Nuclear Fuels Corporation (ANF). This profile describes the status of ANF one year later, principally from the European perspective but with some mention of ANF's worldwide operations to provide a balanced picture. After one year of operation as an affiliate of Siemens/KWU, ANF's role remains as an independent international supplier of nuclear fuel and services to utilities in Europe, the USA and the Far East, but with substantially augmented capabilities resulting from the new affiliation

Collective processing device for spent fuel

International Nuclear Information System (INIS)

Irie, Hiroaki; Taniguchi, Noboru.

1996-01-01

The device of the present invention comprises a sealing vessel, a transporting device for transporting spent fuels to the sealing vessel, a laser beam cutting device for cutting the transported spent fuels, a dissolving device for dissolving the cut spent fuels, and a recovering device for recovering radioactive materials from the spent fuels during processing. Reprocessing treatments comprising each processing of dismantling, shearing and dissolving are conducted in the sealing vessel can ensure a sealing barrier for the radioactive materials (fissionable products and heavy nuclides). Then, since spent fuels can be processed in a state of assemblies, and the spent fuels are easily placed in the sealing vessel, operation efficiency is improved, as well as operation cost is saved. Further, since the spent fuels can be cut by a remote laser beam operation, there can be prevented operator's exposure due to radioactive materials released from the spent fuels during cutting operation. (T.M.)

Natural circulation analysis for the advanced neutron source reactor refueling process 11

Energy Technology Data Exchange (ETDEWEB)

Tucker, R.F.; Dasardhi, S.; Elkassabgi, Y. [Texas A& M Univ., Kingsville, TX (United States); Yoder, G.L. [Oak Ridge National Lab., TN (United States)

1995-09-01

During the refueling process of the Advanced Neutron Source Reactor (ANSR), the spent fuel elements must be moved from the primary coolant loop (containing D{sub 2}O), through a heavy water pool, and finally into a light water spent fuel storage area. The present refueling scheme utilizes remote refueling equipment to move the spent fuel elements through a D{sub 2}O filled stack and tunnel into a temporary storage canal. A transfer lock is used to move the spent fuel elements from the D{sub 2}O-filled interim storage canal to a light water pool. Each spent fuel element must be cooled during this process, using either natural circulation or forced convection. This paper presents a summary of the numerical techniques used to analyze natural circulation cooling of the ANSR fuel elements as well as selected results of the calculations. Details of the analysis indicate that coolant velocities below 10 cm/s exist in the coolant channels under single phase natural circulation conditions. Also, boiling does not occur within the channels if power levels are below a few hundred kW when the core transitions to natural circulation conditions.

Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

International Nuclear Information System (INIS)

Lyon, W.F.; Baker, R.B.; Leggett, R.D.

1991-01-01

In this paper, the conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design is performed. As a first step, an intensive literature survey is completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins is designed and analyzed using the SIEX computer code. The analysis predicts that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors

Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

International Nuclear Information System (INIS)

Lyon, W.F.; Baker, R.B.; Leggett, R.D.

1990-11-01

The conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design has been performed. As a first step, an intensive literature survey was completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins were designed and analyzed using the SIEX computer code. The analysis predicted that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors. 13 refs., 10 figs., 2 tabs

Uranium requirements for advanced fuel cycles in expanding nuclear power systems

International Nuclear Information System (INIS)

Banerjee, S.; Tamm, H.

1978-01-01

When considering advanced fuel cycle strategies in rapidly expanding nuclear power systems, equilibrium analyses do not apply. A computer simulation that accounts for system delay times and fissile inventories has been used to study the effects of different fuel cycles and different power growth rates on uranium consumption. The results show that for a given expansion rate of installed capacity, the main factors that affect resource requirements are the fissile inventory needed to introduce the advanced fuel cycle and the conversion (or breeding) ratio. In rapidly expanding systems, the effect of fissile inventory dominates, whereas in slowly expanding systems, conversion or breeding ratio dominates. Heavy-water-moderated and -cooled reactors, with their high conversion ratios, appear to be adaptable vehicles for accommodating fuel cycles covering a wide range of initial fissile inventories. They are therefore particularly suitable for conserving uranium over a wide range of nuclear power system expansion rates

Finite element analysis of advanced neutron source fuel plates

International Nuclear Information System (INIS)

Luttrell, C.R.

1995-08-01

The proposed design for the Advanced Neutron Source reactor core consists of closely spaced involute fuel plates. Coolant flows between the plates at high velocities. It is vital that adjacent plates do not come in contact and that the coolant channels between the plates remain open. Several scenarios that could result in problems with the fuel plates are studied. Finite element analyses are performed on fuel plates under pressure from the coolant flowing between the plates at a high velocity, under pressure because of a partial flow blockage in one of the channels, and with different temperature profiles

Mixed U/Pu oxide fuel fabrication facility co-processed feed, pelletized fuel

International Nuclear Information System (INIS)

1978-09-01

Two conceptual MOX fuel fabrication facilities are discussed in this study. The first facility in the main body of the report is for the fabrication of LWR uranium dioxide - plutonium dioxide (MOX) fuel using co-processed feed. The second facility in the addendum is for the fabrication of co-processed MOX fuel spiked with 60 Co. Both facilities produce pellet fuel. The spiked facility uses the same basic fabrication process as the conventional MOX plant but the fuel feed incorporates a high energy gamma emitter as a safeguard measure against diversion; additional shielding is added to protect personnel from radiation exposure, all operations are automated and remote, and normal maintenance is performed remotely. The report describes the fuel fabrication process and plant layout including scrap and waste processing; and maintenance, ventilation and safety measures

Research on Elemental Technology of Advanced Nuclear Fuel Performance Verification

International Nuclear Information System (INIS)

Kim, Yong Soo; Lee, Dong Uk; Jean, Sang Hwan; Koo, Min

2003-04-01

Most of current properties models and fuel performance models used in the performance evaluation codes are based on the in-pile data up to 33,000 MWd/MtU. Therefore, international experts are investigating the properties changes and developing advanced prediction models for high burn-up application. Current research is to develop high burn-up fission gas release model for the code and to support the code development activities by collecting data and models, reviewing/assessing the data and models together, and benchmarking the selected models against the appropriate in-pile data. For high burn-up applications, two stage two step fission gas release model is developed based on the real two diffusion process in the grain lattice and grain boundaries of the fission gases and the observation of accelerated release rate in the high burn-up. It is found that the prediction of this model is in excellent agreement with the in-pile measurement results, not only in the low burn-up but also in the high burn-up. This research is found that the importance of thermal conductivity of oxide fuel, especially in the high burn-up, is focused again. It is found that even the temperature dependent models differ from one to another and most of them overestimate the conductivity in the high burn-up. An in-pile data benchmarking of high LHGR fuel rod shows that the difference can reach 30%∼40%, which predicts 400 .deg. C lower than the real fuel centerline temperature. Recent models on the thermal expansion and heat capacity of oxide fuel are found to be well-defined. Irradiation swelling of the oxide fuel are now well-understood that in most cases in LWRs solid fission product swelling is dominant. Thus, the accumulation of in-pile data can enhance the accuracy of the model prediction, rather than theoretical modeling works. Thermo-physical properties of Zircaloy cladding are also well-defined and well-understood except the thermal expansion. However, it turns out that even the

Econometric comparisons of liquid rocket engines for dual-fuel advanced earth-to-orbit shuttles

Science.gov (United States)

Martin, J. A.

1978-01-01

Econometric analyses of advanced Earth-to-orbit vehicles indicate that there are economic benefits from development of new vehicles beyond the space shuttle as traffic increases. Vehicle studies indicate the advantage of the dual-fuel propulsion in single-stage vehicles. This paper shows the economic effect of incorporating dual-fuel propulsion in advanced vehicles. Several dual-fuel propulsion systems are compared to a baseline hydrogen and oxygen system.

Fusion reactor fuel processing

International Nuclear Information System (INIS)

Johnson, E.F.

1972-06-01

For thermonuclear power reactors based on the continuous fusion of deuterium and tritium the principal fuel processing problems occur in maintaining desired compositions in the primary fuel cycled through the reactor, in the recovery of tritium bred in the blanket surrounding the reactor, and in the prevention of tritium loss to the environment. Since all fuel recycled through the reactor must be cooled to cryogenic conditions for reinjection into the reactor, cryogenic fractional distillation is a likely process for controlling the primary fuel stream composition. Another practical possibility is the permeation of the hydrogen isotopes through thin metal membranes. The removal of tritium from the ash discharged from the power system would be accomplished by chemical procedures to assure physiologically safe concentration levels. The recovery process for tritium from the breeder blanket depends on the nature of the blanket fluids. For molten lithium the only practicable possibility appears to be permeation from the liquid phase. For molten salts the process would involve stripping with inert gas followed by chemical recovery. In either case extremely low concentrations of tritium in the melts would be desirable to maintain low tritium inventories, and to minimize escape of tritium through unwanted permeation, and to avoid embrittlement of metal walls. 21 refs

Advanced Reactor Fuels Irradiation Experiment Design Objectives

Energy Technology Data Exchange (ETDEWEB)

Chichester, Heather Jean MacLean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven Lowe [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dempsey, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

Advanced Reactor Fuels Irradiation Experiment Design Objectives

International Nuclear Information System (INIS)

Chichester, Heather Jean MacLean; Hayes, Steven Lowe; Dempsey, Douglas; Harp, Jason Michael

2016-01-01

This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

Fuel, structural material and coolant for an advanced fast micro-reactor

International Nuclear Information System (INIS)

Nascimento, Jamil A. do; Guimaraes, Lamartine N.F.; Ono, Shizuca

2011-01-01

The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials. (author)

Alternative fuels and advanced technology vehicles : issues in Congress

Science.gov (United States)

2009-02-13

Alternative fuels and advanced technology vehicles are seen by proponents as integral to improving urban air quality, decreasing dependence on foreign oil, and reducing emissions of greenhouse gases. However, major barriers especially economics curre...

Development of an advanced 16x165 Westinghouse type PWR fuel assembly for Slovenia

International Nuclear Information System (INIS)

Boone, M. L.; King, S. J.; Pulver, E. F.; Jeon, K.-L.; Esteves, R.; Kurincic, B.

2004-01-01

Industrias Nucleares do Brasil (INB), KEPCO Nuclear Fuel Company, Ltd. (KNFC), and Westinghouse Electric Company (Westinghouse) have jointly designed an advanced 16x16 Westinghouse type PWR fuel assembly. This advanced 16x16 Westinghouse type PWR fuel assembly, which will be implemented in both Kori Unit 2 (in Korea) and Angra Unit 1 (in Brazil) in January and March 2005, respectively, is an integral part of the utilities fuel management strategy. This same fuel design has also been developed for future use in Krsko Unit 1 (in Slovenia). In this paper we will describe the front-end nuclear fuel management activities utilized by the joint development team and describe how these activities played an integral part in defining the direction of the advanced 16x16 Westinghouse type PWR fuel assembly design. Additionally, this paper will describe how this design demonstrates improved margins under high duty plant operating conditions. The major reason for initiating this joint development program was to update the current 16x16 fuel assembly, which is also called 16STD. The current 16STD fuel assembly contains a non-optimized fuel rod diameter for the fuel rod pitch (i.e. 9.5 mm OD fuel rods at a 0.485 inch pitch), non-neutronic efficient components (i.e. Inconel Mid grids), no Intermediate Flow Mixer (IFM) grids, and other mechanical features. The advanced 16x16 fuel assembly is being designed for peak rod average burnups of up to 75 MWd/kgU and will use an optimized fuel rod diameter (i.e. 9.14 mm OD ZIRLO TM fuel rods), neutronic efficient components (i.e. ZIRLO TM Mid grids), ZIRLO TM Intermediate Flow Mixer (IFM) grids to improve Departure from Nucleate Boiling (DNB) margin, and many other mechanical features that improve design margins. Nuclear design activities in the areas of fuel cycle cost and fuel management were performed in parallel to the fuel assembly design efforts. As the change in reactivity due to the change in the fuel rod diameter influences directly

Japan-IAEA Workshops on Advanced Safeguards for Future Nuclear Fuel Cycles

International Nuclear Information System (INIS)

Hoffheins, B.; Hori, M.; Suzuki, M.; Kuno, Y.; Kimura, N.; Naito, K.; Hosoya, M.; Khlebnikov, N.; Whichello, J.; Zendel, M.

2010-01-01

Beginning in 2007, the Japan Atomic Energy Agency (JAEA) and the International Atomic Energy Agency (IAEA) Department of Safeguards initiated a workshop series focused on advanced safeguards technologies for the future nuclear fuel cycle (NFC). The goals for these workshops were to address safeguards challenges, to share implementation experiences, to discuss fuel cycle plans and promising research and development, and to address other issues associated with safeguarding new fuel cycle facilities. Concurrently, the workshops also served to promote dialog and problem solving, and to foster closer collaborations for facility design and planning. These workshops have sought participation from IAEA Member States' support programmes (MSSP), the nuclear industry, R and D organizations, state systems of accounting and control (SSAC), regulators and inspectorates to ensure that all possible stakeholder views can be shared in an open process. Workshop presentations have covered, inter alia, national fuel cycle programs and plans, research progress in proliferation resistance (PR) and safeguardability, approaches for nuclear measurement accountancy of large material throughputs and difficult to access material, new and novel radiation detectors with increased sensitivity and automation, and lessons learned from recent development and operation of safeguards systems for complex facilities and the experiences of integrated safeguards (IS) in Japan. Although the title of the workshops presumes an emphasis on technology, participants recognized that early planning and organization, coupled with close cooperation among stakeholders, that is, through the application of 'Safeguards by Design' (SBD) processes that include nuclear safety and security coordination, 'Remote Inspections' and 'Joint-Use of Equipment (JUE)' would be required to enable more successful implementations of safeguards at future NFC facilities. The needs to cultivate the future workforce, effectively preserve

Development of An Advanced JP-8 Fuel

Science.gov (United States)

1993-12-01

included the Microthermal Precipitation Test (MTP), Fuel Reactor Test, Hot Liquid Process Simulator (HLPS), and Isothermal Corrosion Oxidation Test (ICOT... Microthermal Precipitation Test The impetus for this development effort was the need for a screening test that could discriminate between fuels of...varying propensity to produce thermally induced insoluble particulate material in the bulk fuel. The Microthermal Precipitation (MTP) test thermally

Detailed modeling of common rail fuel injection process

NARCIS (Netherlands)

Seykens, X.L.J.; Somers, L.M.T.; Baert, R.S.G.

2005-01-01

Modeling of fuel injection equipment is a tool that is used increasingly for explaining or predicting the effect of advanced diesel injection strategies on combustion and emissions. This paper reports on the modeling of the high-pressure part of a research type Heavy Duty Common Rail (CR) fuel

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.

TALSPEAK Chemistry in Advanced Nuclear Fuel Cycles

International Nuclear Information System (INIS)

Nilsson, Mikael; Nash, Kenneth L.

2008-01-01

The separation of trivalent transplutonium actinides from fission product lanthanide ions represents a challenging aspect of advanced nuclear fuel partitioning schemes. The challenge of this separation could be amplified in the context of the AFCI-UREX+1a process, as Np and Pu will accompany the minor actinides to this stage of separation. At present, the baseline lanthanide-actinide separation method is the TALSPEAK (Trivalent Actinide - Lanthanide Separation by Phosphorus reagent Extraction from Aqueous complexes) process. TALSPEAK was developed in the late 1960's at Oak Ridge National Laboratory and has been demonstrated at pilot scale. This process relies on the complex interaction between an organic and an aqueous phase both containing complexants for selectively separating the trivalent actinide. The 3 complexing components are: the di(2-ethylhexyl) phosphoric acid (HDEHP), the lactic acid (HL) and the diethylenetriamine-N,N,N',N'',N''-pentaacetic acid (DTPA). In this report we discuss observations on kinetic and thermodynamic features described in the prior literature and describe some results of our ongoing research on basic chemical features of this system. The information presented indicates that the lactic acid buffer participates in the net operation of the TALSPEAK process in a manner that is not explained by existing information on the thermodynamic features if the known Eu(III)-lactate species. (authors)

Alternative Fuel and Advanced Technology Commercial Lawn Equipment (Brochure)

Energy Technology Data Exchange (ETDEWEB)

2014-10-01

The U.S. Department of Energy's Clean Cities program produced this guide to help inform the commercial mowing industry about product options and potential benefits. This guide provides information about equipment powered by propane, ethanol, compressed natural gas, biodiesel, and electricity, as well as advanced engine technology. In addition to providing an overview for organizations considering alternative fuel lawn equipment, this guide may also be helpful for organizations that want to consider using additional alternative fueled equipment.

Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

International Nuclear Information System (INIS)

Mella, R.; Wenman, M.R.

2013-01-01

Thermo-mechanical contributions to pellet–clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS’s well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used

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

International Nuclear Information System (INIS)

Wilson, Paul

2009-01-01

Develops the tools to investigate 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.

Mathematical modeling of biomass fuels formation process

International Nuclear Information System (INIS)

Gaska, Krzysztof; Wandrasz, Andrzej J.

2008-01-01

The increasing demand for thermal and electric energy in many branches of industry and municipal management accounts for a drastic diminishing of natural resources (fossil fuels). Meanwhile, in numerous technical processes, a huge mass of wastes is produced. A segregated and converted combustible fraction of the wastes, with relatively high calorific value, may be used as a component of formed fuels. The utilization of the formed fuel components from segregated groups of waste in associated processes of co-combustion with conventional fuels causes significant savings resulting from partial replacement of fossil fuels, and reduction of environmental pollution resulting directly from the limitation of waste migration to the environment (soil, atmospheric air, surface and underground water). The realization of technological processes with the utilization of formed fuel in associated thermal systems should be qualified by technical criteria, which means that elementary processes as well as factors of sustainable development, from a global viewpoint, must not be disturbed. The utilization of post-process waste should be preceded by detailed technical, ecological and economic analyses. In order to optimize the mixing process of fuel components, a mathematical model of the forming process was created. The model is defined as a group of data structures which uniquely identify a real process and conversion of this data in algorithms based on a problem of linear programming. The paper also presents the optimization of parameters in the process of forming fuels using a modified simplex algorithm with a polynomial worktime. This model is a datum-point in the numerical modeling of real processes, allowing a precise determination of the optimal elementary composition of formed fuels components, with assumed constraints and decision variables of the task

Compound process fuel cycle concept

International Nuclear Information System (INIS)

Ikegami, Tetsuo

2005-01-01

Mass flow of light water reactor spent fuel for a newly proposed nuclear fuel cycle concept 'Compound Process Fuel Cycle' has been studied in order to assess the capacity of the concept for accepting light water reactor spent fuels, taking an example for boiling water reactor mixed oxide spent fuel of 60 GWd/t burn-up and for a fast reactor core of 3 GW thermal output. The acceptable heavy metal of boiling water reactor mixed oxide spent fuel is about 3.7 t/y/reactor while the burn-up of the recycled fuel is about 160 GWd/t and about 1.6 t/y reactor with the recycled fuel burn-up of about 300 GWd/t, in the case of 2 times recycle and 4 times recycle respectively. The compound process fuel cycle concept has such flexibility that it can accept so much light water reactor spent fuels as to suppress the light water reactor spent fuel pile-up if not so high fuel burn-up is expected, and can aim at high fuel burn-up if the light water reactor spent fuel pile-up is not so much. Following distinctive features of the concept have also been revealed. A sort of ideal utilization of boiling water reactor mixed oxide spent fuel might be achieved through this concept, since both plutonium and minor actinide reach equilibrium state beyond 2 times recycle. Changes of the reactivity coefficients during recycles are mild, giving roughly same level of reactivity coefficients as the conventional large scale fast breeder core. Both the radio-activity and the heat generation after 4 year cooling and after 4 times recycle are less than 2.5 times of those of the pre recycle fuel. (author)

Experiences and Trends of Manufacturing Technology of Advanced Nuclear Fuels

International Nuclear Information System (INIS)

2012-08-01

The 'Atoms for Peace' mission initiated in the mid-1950s paved the way for the development and deployment of nuclear fission reactors as a source of heat energy for electricity generation in nuclear power reactors and as a source of neutrons in non-power reactors for research, materials irradiation, and testing and production of radioisotopes. The fuels for nuclear reactors are manufactured from natural uranium (∼99.3% 238 U + ∼0.7% 235 U) and natural thorium (∼100% 232 Th) resources. Currently, most power and research reactors use 235 U, the only fissile isotope found in nature, as fuel. The fertile isotopes 238 U and 232 Th are transmuted in the reactor to human-made 239 Pu and 233 U fissile isotopes, respectively. Likewise, minor actinides (MA) (Np, Am and Cm) and other plutonium isotopes are also formed by a series of neutron capture reactions with 238 U and 235 U. Long term sustainability of nuclear power will depend to a great extent on the efficient, safe and secure utilization of fissile and fertile materials. Light water reactors (LWRs) account for more than 82% of the operating reactors, followed by pressurized heavy water reactors (PHWRs), which constitute ∼10% of reactors. LWRs will continue to dominate the nuclear power market for several decades, as long as economically viable natural uranium resources are available. Currently, the plutonium obtained from spent nuclear fuel is subjected to mono recycling in LWRs as uranium-plutonium mixed oxide (MOX), containing up to 12% PuO 2 , in a very limited way. The reprocessed uranium (RepU) is also re-enriched and recycled in LWRs in a few countries. Unfortunately, the utilization of natural uranium resources in thermal neutron reactors is 2 and MOX fuel technology has matured during the past five decades. These fuels are now being manufactured, used and reprocessed on an industrial scale. Mixed uranium- plutonium monocarbide (MC), mononitride (MN) and U-Pu-Zr alloys are recognized as advanced fuels

Alternative Fuel and Advanced Technology Commercial Lawn Equipment (Spanish version); Clean Cities, Energy Efficiency & Renewable Energy (EERE)

Energy Technology Data Exchange (ETDEWEB)

Nelson, Erik

2015-06-01

Powering commercial lawn equipment with alternative fuels or advanced engine technology is an effective way to reduce U.S. dependence on petroleum, reduce harmful emissions, and lessen the environmental impacts of commercial lawn mowing. Numerous alternative fuel and fuel-efficient advanced technology mowers are available. Owners turn to these mowers because they may save on fuel and maintenance costs, extend mower life, reduce fuel spillage and fuel theft, and demonstrate their commitment to sustainability.

Fuels Processing Laboratory

Data.gov (United States)

Federal Laboratory Consortium — NETL’s Fuels Processing Laboratory in Morgantown, WV, provides researchers with the equipment they need to thoroughly explore the catalytic issues associated with...

Design and analysis of CANDU advanced fuel -Development of the advanced CANDU technology-

International Nuclear Information System (INIS)

Seok, Ho Cheon; Shim, Ki Seop; Byeon, Taek Sang; Park, Kwang Seok; Kim, Bong Ki; Lee, Yeong Uk; Jeong, Chang Joon; Oh, Deok Joo; Lee, Ui Joo; Park, Joo Hwan; Lee, Sang Yong; Jeong, Beop Dong; Choi, Han Rim; Lee, Yeong Jin; Choi, Cheol Jin; Choi, Jong Ho; Lee, Kwang Won; Cho, Cheon Hyi; On, Myeong Ryong; Kim, Taek Mo; Lim, Hong Sik; Lee, Kang Moon; Lee, Nam Ho; Lee, Kyu Hyeong

1994-07-01

It has been projected that a total of 5 pressurized heavy water reactors (PHWR) including Wolsong 1 under operation and Wolsong 2, 3 and 4 under construction will be operated by 2006, and so about 500 ton of natural uranium will be consumed every year and a lot of spent fuels will be generated. Therefore, the ultimate goal of this R and D project is to develop the CANDU advanced fuel having the following capabilities compared with existing standard fuel: (1) To reduce linear heat generation rating by more than 15% (i.e., less than 50 kW/m), (2) To extend fuel burnup by more than 3 times (i.e., higher than 21,000 MWD/MTU), and (3) To increase critical channel power by more than 5%. In accordance, the followings are performed in this fiscal year: (1) Undertake CANFLEX-NU design and thermalmechanical performance analysis, and prepare design documents, (2) Establish reactor physics analysis code system, and investigate the compativility of the CANFLEX-NU fuel with the standard 37-element fuel in the CANDU-6 reactor. (3) Establish safety analysis methodology with the assumption of the CANFLEX-NU loaded CANDU-6 reactor, and perform the preliminary thermalhydraulic and fuel behavior for the selected DBA accidents, (4) Investigate reactor physics analysis code system as pre-study for CANFLEX-SEU loaded reactors

Development of Head-end Pyrochemical Reduction Process for Advanced Oxide Fuels

International Nuclear Information System (INIS)

Park, B. H.; Seo, C. S.; Hur, J. M.; Jeong, S. M.; Hong, S. S.; Choi, I. K.; Choung, W. M.; Kwon, K. C.; Lee, I. W.

2008-12-01

The development of an electrolytic reduction technology for spent fuels in the form of oxide is of essence to introduce LWR SFs to a pyroprocessing. In this research, the technology was investigated to scale a reactor up, the electrochemical behaviors of FPs were studied to understand the process and a reaction rate data by using U 3 O 8 was obtained with a bench scale reactor. In a scale of 20 kgHM/batch reactor, U 3 O 8 and Simfuel were successfully reduced into metals. Electrochemical characteristics of LiBr, LiI and Li 2 Se were measured in a bench scale reactor and an electrolytic reduction cell was modeled by a computational tool

Method for pre-processing LWR spent fuel

International Nuclear Information System (INIS)

Otsuka, Katsuyuki; Ebihara, Hikoe.

1986-01-01

Purpose: To facilitate the decladding of spent fuel, cladding tube processing, and waste gas recovery, and to enable the efficient execution of main re-processing process thereafter. Constitution: Spent fuel assemblies are sent to a cutting process where they are cut into chips of easy-to-process size. The chips, in a thermal decladding process, undergo a thermal cycle processing in air with the processing temperatures increased and decreased within the range of from 700 deg C to 1200 deg C, oxidizing zircaloy comprising the cladding tubes into zirconia. The oxidized cladding tubes have a number of fine cracks and become very brittle and easy to loosen off from fuel pellets when even a slight mechanical force is applied thereto, thus changing into a form of powder. Processed products are then separated into zirconia sand and fuel pellets by a gravitational selection method or by a sifting method, the zirconia sand being sent to a waste processing process and the fuel pellets to a melting-refining process. (Yoshino, Y.)

Advanced fuels for gas turbines: Fuel system corrosion, hot path deposit formation and emissions

International Nuclear Information System (INIS)

Seljak, Tine; Širok, Brane; Katrašnik, Tomaž

2016-01-01

Highlights: • Technical feasibility analysis of alternative fuels requires a holistic approach. • Fuel, combustion, corrosion and component functionality are strongly related. • Used approach defines design constraints for microturbines using alternative fuels. - Abstract: To further expand the knowledge base on the use of innovative fuels in the micro gas turbines, this paper provides insight into interrelation between specific fuel properties and their impact on combustion and emission formation phenomena in micro gas turbines for stationary power generation as well as their impact on material corrosion and deposit formation. The objective of this study is to identify potential issues that can be related to specific fuel properties and to propose counter measures for achieving stable, durable, efficient and low emission operation of the micro gas turbine while utilizing advanced/innovative fuels. This is done by coupling combustion and emission formation analyses to analyses of material degradation and degradation of component functionality while interpreting them through fuel-specific properties. To ensure sufficiently broad range of fuel properties to demonstrate the applicability of the method, two different fuels with significantly different properties are analysed, i.e. tire pyrolysis oil and liquefied wood. It is shown that extent of required micro gas turbine adaptations strongly correlates with deviations of the fuel properties from those of the baseline fuel. Through the study, these adaptations are supported by in-depth analyses of impacts of fuel properties on different components, parameters and subsystems and their quantification. This holistic approach is further used to propose methodologies and innovative approaches for constraining a design space of micro gas turbine to successfully utilize wide spectra of alternative/innovative fuels.

Review of low pressure plasma processing of proton exchange membrane fuel cell electrocatalysts

OpenAIRE

Brault , Pascal

2016-01-01

Review article; International audience; The present review is describing recent advances in plasma deposition and treatment of low temperature proton exchange membrane fuel cells electrocatalysts. Interest of plasma processing for growth of platinum based, non-precious and metal free electrocatalysts is highlighted. Electrocatalysts properties are tentatively correlated to plasma parameters.

Fuel processing for molten-salt reactors

International Nuclear Information System (INIS)

Hightower, J.R. Jr.

1976-01-01

Research devoted to development of processes for the isolation of protactinium and for the removal of fission products from molten-salt breeder reactors is reported. During this report period, engineering development progressed on continuous fluorinators for uranium removal, the metal transfer process for rare-earth removal, the fuel reconstitution step, and molten salt--bismuth contactors to be used in reductive extraction processes. The metal transfer experiment MTE-3B was started. In this experiment all parts of the metal transfer process for rare-earth removal are demonstrated using salt flow rates which are about 1 percent of those required to process the fuel salt in a 1000-MW(e) MSBR. During this report period the salt and bismuth phases were transferred to the experimental vessels, and two runs with agitator speeds of 5 rps were made to measure the rate of transfer of neodymium from the fluoride salt to the Bi--Li stripper solution. The uranium removed from the fuel salt by fluorination must be returned to the processed salt in the fuel reconstitution step before the fuel salt is returned to the reactor. An engineering experiment to demonstrate the fuel reconstitution step is being installed. In this experiment gold-lined equipment will be used to avoid introducing products of corrosion by UF 6 and UF 5 . Alternative methods for providing the gold lining include electroplating and mechanical fabrication

Radioactive waste management and advanced nuclear fuel cycle technologies

International Nuclear Information System (INIS)

2007-01-01

In 2007 ENEA's Department of Nuclear Fusion and Fission, and Related Technologies acted according to national policy and the role assigned to ENEA FPN by Law 257/2003 regarding radioactive waste management and advanced nuclear fuel cycle technologies

Modified ADS molten salt processes for back-end fuel cycle of PWR spent fuel

International Nuclear Information System (INIS)

Choi, In-Kyu; Yeon, Jei-Won; Kim, Won-Ho

2002-01-01

The back-end fuel cycle concept for PWR spent fuel is explained. This concept is adequate for Korea, which has operated both PWR and CANDU reactors. Molten salt processes for accelerator driven system (ADS) were modified both for the transmutation of long-lived radioisotopes and for the utilisation of the remained fissile uranium in PWR spent fuels. Prior to applying molten salt processes to PWR fuel, hydrofluorination and fluorination processes are applied to obtain uranium hexafluoride from the spent fuel pellet. It is converted to uranium dioxide and fabricated into CANDU fuel. From the remained fluoride compounds, transuranium elements can be separated by the molten salt technology such as electrowinning and reductive extraction processes for transmutation purpose without weakening the proliferation resistance of molten salt technology. The proposed fuel cycle concept using fluorination processes is thought to be adequate for our nuclear program and can replace DUPIC (Direct Use of spent PWR fuel in CANDU reactor) fuel cycle. Each process for the proposed fuel cycle concept was evaluated in detail

The PBMR fuel plant: Proven technology in an advanced safety environment

International Nuclear Information System (INIS)

Braehler, G.; Froschauer, K.; Welbers, P.; Boyes, D.

2008-01-01

The PBMR Fuel Plant (PFP), to be constructed at the Pelindaba site near Johannesburg will fuel the first South African Pebble Bed Modular Reactor. The qualification of the PBMR fuel shall be based on past experience with fuel which was produced in the German NUKEM/HOBEG plant and irradiated in the German AVR reactor. Accordingly, the PFP must produce the same fuel as the German plant did, and consequently, the design of the PFP has in essence to be a copy of the NUKEM/HOBEG plant. As a reminder this plant had been operated in accordance with the German regulatory rules which were defined in the years 1970/80. Since then, the requirements with regard to radiological protection, criticality safety and emission control have been significantly tightened, and of course the PFP must be designed in accordance with the most advanced international norms and standards. The implications which follow from these two potentially conflicting requirements, as defined above, are highlighted, and technical solutions are presented. Hence, the change from administrative criticality safety control to technical control, i.e. the application of safe geometry as far as possible. and the introduction of technical solutions for the remaining safe mass regime will be described. A lot of equipment in the Kernel area and in the recycling areas needed to be redesigned in safe geometry. The sensitive processes for Kernel Calcining, for the Coating and the Over-coating remain under safe mass regime, but the safety against criticality is completely independent from staff activities and based on technical measures. A new concept for safe storage of large volumes of Uranium-containing liquids has been developed. Also, the change from relatively open handling of Uranium to the application of containment enclosures wherever release of radioactivity into the room atmosphere is possible, will be addressed. This change required redesign of all process steps requiring the handling of dry Uranium oxides

A secondary fuel removal process: plasma processing

Energy Technology Data Exchange (ETDEWEB)

Min, J Y; Kim, Y S [Hanyang Univ., Seoul (Korea, Republic of); Bae, K K; Yang, M S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-07-01

Plasma etching process of UO{sub 2} by using fluorine containing gas plasma is studied as a secondary fuel removal process for DUPIC (Direct Use of PWR spent fuel Into Candu) process which is taken into consideration for potential future fuel cycle in Korea. CF{sub 4}/O{sub 2} gas mixture is chosen for reactant gas and the etching rates of UO{sub 2} by the gas plasma are investigated as functions of CF{sub 4}/O{sub 2} ratio, plasma power, substrate temperature, and plasma gas pressure. It is found that the optimum CF{sub 4}/O{sub 2} ratio is around 4:1 at all temperatures up to 400 deg C and the etching rate increases with increasing r.f. power and substrate temperature. Under 150W r.f. power the etching rate reaches 1100 monolayers/min at 400 deg C, which is equivalent to about 0.5mm/min. (author).

The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

International Nuclear Information System (INIS)

Grover, S. Blaine

2009-01-01

The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

Design concepts and advanced manipulator development for nuclear fuel cycle facilities

International Nuclear Information System (INIS)

Feldman, M.J.

1985-01-01

In the Fuel Recycle Division, Consolidated Fuel Reprocessing Program at the Oak Ridge National Laboratory, a comprehensive remote systems development program has existed for the past seven years. The new remote technology under development is expected to significantly improve remote operations by extending the range of tasks accomplished by remote means and increasing the efficiency of remote work undertaken. The application of advanced manipulation is viewed as an essential part of a series of design directions whose sum describes a somewhat unique blend of old and new technology. A design direction based upon the Teletec concept is explained and recent progress in the development of an advanced servomanipulator-based maintenance concept is summarized to show that a new generation of remote systems is feasible through advanced technology. 14 refs., 14 figs

Study on advanced nuclear fuel cycle of PWR/CANDU synergism

International Nuclear Information System (INIS)

Xie Zhongsheng; Huo Xiaodong

2002-01-01

According to the concrete condition that China has both PWR and CANDU reactors, one of the advanced nuclear fuel cycle strategy of PWR/CANDU synergism ws proposed, i.e. the reprocessed uranium of spent PWR fuel was used in CANDU reactor, which will save the uranium resource, increase the energy output, decrease the quantity of spent fuels to be disposed and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, the transition from the natural uranium to the recycled uranium (RU) can be completed without any changes of the structure of reactor core and operation mode. Furthermore, because of the low radiation level of RU, which is acceptable for CANDU reactor fuel fabrication, the present product line of fuel elements of CANDU reactor only need to be shielded slightly, also the conditions of transportation, operation and fuel management need not to be changed. Thus this strategy has significant practical and economical benefit

High-Level Functional and Operational Requirements for the Advanced Fuel Cycle Facility

International Nuclear Information System (INIS)

Charles Park

2006-01-01

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

STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

Energy Technology Data Exchange (ETDEWEB)

Davenport, Michael; Petti, D. A.; Palmer, Joe

2016-11-01

The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

Design study on advanced nuclear fuel recycling system by pyrometallurgical reprocessing technology

Energy Technology Data Exchange (ETDEWEB)

Kasai, Yoshimitsu; Kakehi, Isao; Moro, Satoshi; Tobe, Kenji; Kawamura, Fumio; Higashi, Tatsuhiro; Yonezawa, Shigeaki [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Yoshiuji, Takahiro

1998-12-01

The Japan Nuclear Fuel Cycle Development Institute is conducting research and development on the nuclear fuel recycling system, which will improve the economy, safety, and environmental impact of the nuclear fuel recycling system in the age of the FBR. The System Engineering Division in the O-arai Engineering Center has conducted a design study on an advanced nuclear fuel recycling system for FBRs by using pyrometallurgical reprocessing technology. The system is an economical and compact module-type system, and can be used for reprocessing oxide fuel and also new types of fuel (metal fuel and nitride fuel). This report describes the concept of this system and results of the design study. (author)

Design study on advanced nuclear fuel recycling system by pyrometallurgical reprocessing technology

International Nuclear Information System (INIS)

Kasai, Yoshimitsu; Kakehi, Isao; Moro, Satoshi; Tobe, Kenji; Kawamura, Fumio; Higashi, Tatsuhiro; Yonezawa, Shigeaki; Yoshiuji, Takahiro

1998-01-01

The Japan Nuclear Fuel Cycle Development Institute is conducting research and development on the nuclear fuel recycling system, which will improve the economy, safety, and environmental impact of the nuclear fuel recycling system in the age of the FBR. The System Engineering Division in the O-arai Engineering Center has conducted a design study on an advanced nuclear fuel recycling system for FBRs by using pyrometallurgical reprocessing technology. The system is an economical and compact module-type system, and can be used for reprocessing oxide fuel and also new types of fuel (metal fuel and nitride fuel). This report describes the concept of this system and results of the design study. (author)

THE TESTING OF COMMERCIALLY AVAILABLE ENGINEERING AND PLANT SCALE ANNULAR CENTRIFUGAL CONTACTORS FOR THE PROCESSING OF SPENT NUCLEAR FUEL

International Nuclear Information System (INIS)

Jack D. Law; David Meikrantz; Troy Garn; Nick Mann; Scott Herbst

2006-01-01

Annular centrifugal contactors are being evaluated for process scale solvent extraction operations in support of United State Advanced Fuel Cycle Initiative goals. These contactors have the potential for high stage efficiency if properly employed and optimized for the application. Commercially available centrifugal contactors are being tested at the Idaho National Laboratory to support this program. Hydraulic performance and mass transfer efficiency have been measured for portions of an advanced nuclear fuel cycle using 5-cm diameter annular centrifugal contactors. Advanced features, including low mix sleeves and clean-in-place rotors, have also been evaluated in 5-cm and 12.5-cm contactors

Design of an advanced human-centered supervisory system for a nuclear fuel reprocessing system

International Nuclear Information System (INIS)

Riera, B.; Lambert, M.; Martel, G.

1999-01-01

In the field of highly automated processes, our research concerns supervisory system design adapted to supervisory and default diagnosis by human operators. The interpretation of decisional human behaviour models shows that the tasks of human operators require different information, which has repercussions on the supervisory system design. We propose an advanced human-centred supervisory system (AHCSS) which is more adapted to human-beings, because it integrates new representation of the production system,(such as functional and behavioural aspects) with the use of advanced algorithms of detection and location. Based on an approach using these new concepts, and AHCSS was created for a nuclear fuel reprocessing system. (authors)

ACSEPT, a new step in the future demonstration of advanced fuel processing

International Nuclear Information System (INIS)

Bourg, Stephane; Hill, Clement; Caravaca, Concha; Ekberg, Christian; Rhodes, Chris

2010-01-01

Actinide recycling by separation and transmutation is considered world wide and particularly in several European countries as one of the most promising strategies to reduce the inventory of radioactive waste, thus contributing to making nuclear energy sustainable. By joining together European universities, nuclear research bodies and major industrial players in a multi-disciplinary consortium, the FP7 EURATOM Fission Project ACSEPT provides the sound basis and fundamental improvements for future demonstrations of fuel treatment in strong connection with fuel fabrication techniques. In accordance with the Strategic Research Agenda of the Sustainable Nuclear Energy Technology Platform (SNE-TP), the timelines of this four-year R and D project (2008-2012) should allow the offering of technical solutions in terms of separation process that may be reviewed by governments, European utilities as well as technology providers at that time horizon. By showing a technically feasible recycling of actinides strategy, ACSEPT will certainly produce positive arguments in the sense that European decision makers, and more globally public opinion, could be convinced that technical solutions for a better management of nuclear wastes are now technologically feasible. ACSEPT is thus an essential contribution to the demonstration, in the long term, of the potential benefits of actinide recycling to minimise the burden on geological repositories. To succeed, ACSEPT is organised in three technical domains: i) Considering technically mature aqueous separation processes, ACSEPT will optimise and select the most promising ones dedicated either to actinide partitioning or to group actinide separation. These developments are appropriately balanced with an exploratory research focused on the design of new molecules. ii) Concerning pyrochemical separation processes, ACSEPT first focuses on the enhancement of the two reference cores of process selected within EUROPART. R and D efforts shall also be

Development of Head-end Pyrochemical Reduction Process for Advanced Oxide Fuels

Energy Technology Data Exchange (ETDEWEB)

Park, B. H.; Seo, C. S.; Hur, J. M.; Jeong, S. M.; Hong, S. S.; Choi, I. K.; Choung, W. M.; Kwon, K. C.; Lee, I. W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-12-15

The development of an electrolytic reduction technology for spent fuels in the form of oxide is of essence to introduce LWR SFs to a pyroprocessing. In this research, the technology was investigated to scale a reactor up, the electrochemical behaviors of FPs were studied to understand the process and a reaction rate data by using U{sub 3}O{sub 8} was obtained with a bench scale reactor. In a scale of 20 kgHM/batch reactor, U{sub 3}O{sub 8} and Simfuel were successfully reduced into metals. Electrochemical characteristics of LiBr, LiI and Li{sub 2}Se were measured in a bench scale reactor and an electrolytic reduction cell was modeled by a computational tool.

Development of advanced LWR fuel cladding

Energy Technology Data Exchange (ETDEWEB)

Jeong, Yong Hwan; Park, S. Y.; Lee, M. H. [and others

2000-04-01

This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out.

Development of advanced LWR fuel cladding

International Nuclear Information System (INIS)

Jeong, Yong Hwan; Park, S. Y.; Lee, M. H.

2000-04-01

This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out

Process development and exergy cost sensitivity analysis of a hybrid molten carbonate fuel cell power plant and carbon dioxide capturing process

Science.gov (United States)

Mehrpooya, Mehdi; Ansarinasab, Hojat; Moftakhari Sharifzadeh, Mohammad Mehdi; Rosen, Marc A.

2017-10-01

An integrated power plant with a net electrical power output of 3.71 × 105 kW is developed and investigated. The electrical efficiency of the process is found to be 60.1%. The process includes three main sub-systems: molten carbonate fuel cell system, heat recovery section and cryogenic carbon dioxide capturing process. Conventional and advanced exergoeconomic methods are used for analyzing the process. Advanced exergoeconomic analysis is a comprehensive evaluation tool which combines an exergetic approach with economic analysis procedures. With this method, investment and exergy destruction costs of the process components are divided into endogenous/exogenous and avoidable/unavoidable parts. Results of the conventional exergoeconomic analyses demonstrate that the combustion chamber has the largest exergy destruction rate (182 MW) and cost rate (13,100 /h). Also, the total process cost rate can be decreased by reducing the cost rate of the fuel cell and improving the efficiency of the combustion chamber and heat recovery steam generator. Based on the total avoidable endogenous cost rate, the priority for modification is the heat recovery steam generator, a compressor and a turbine of the power plant, in rank order. A sensitivity analysis is done to investigate the exergoeconomic factor parameters through changing the effective parameter variations.

Recent advances in hardware and software are to improve spent fuel measurements

International Nuclear Information System (INIS)

Staples, P.; Beddingfield, D.H.; Lestone, J.P.; Pelowitz, D.G.; Bytchkov, M.; Starovich, Z.; Harizanov, I.; Luna-Vellejo, J.; Lavender, C.

2001-01-01

Vast quantities of spent fuel are available for safeguard measurements, primarily in Commonwealth of Independent States (CIS) of the former Soviet Union. This spent fuel, much of which consists of long-cooling-time material, is going to become less unique in the world safeguards arena as reprocessing projects or permanent repositories continue to be delayed or postponed. The long cooling time of many of the spent fuel assemblies being prepared for intermediate term storage in the CIS countries promotes the possibility of increased accuracy in spent fuel assays. This improvement is made possible through the process of decay of the Curium isotopes and of fission products. An important point to consider for the future that could advance safeguards measurements for reverification and inspection would be to determine what safeguards requirements should be imposed upon this 'new' class of spent fuel, Improvements in measurement capability will obviously affect the safeguards requirements. What most significantly enables this progress in spent fuel measurements is the improvement in computer processing power and software enhancements leading to user-friendly Graphical User Interfaces (GUT's). The software used for these projects significantly reduces the IAEA inspector's time expenditure for both learning and operating computer and data acquisition systems, At the same time, by standardizing the spent fuel measurements, it is possible to increase reproducibility and reliability of the measurement data. Hardware systems will be described which take advantage of the increased computer control available to enable more complex measurement scenarios. A specific example of this is the active regulation of a spent fuel neutron coincident counter's 3 He tubes high voltage, and subsequent scaling of measurement results to maintain a calibration for direct assay of the plutonium content of Fast Breeder Reactor spent fuel. The plutonium content has been successfully determined for

Process and quality control in manufacturing of nuclear fuel assemblies of LWRs

International Nuclear Information System (INIS)

Dietrich, M.; Hoff, A.; Reimann, P.

2000-01-01

Manufacturing of nuclear fuel assemblies requires a multitude of different process and quality methods to assure and maintain a high quality level. In recent years methods have been applied which prevent deviations rather than detect deviant products. This paper gives an example on how to control a complex manufacturing process by using a small number of key parameters and second, it demonstrates the importance of graphical data evaluation and presentation methods. In the past many product and product characteristics were inspected m comparison with specification limits only. However, todays methods allow the early identification of trends, increase of variation, shifts disturbances etc. before the product characteristics exceed the specification limits. These methods are process control charts, x-y-plots, boxplots, failure mode and effect analysis (FMEA), process capability numbers and others. This paper demonstrates the beneficial use of some of the methods by presenting selected examples applied at Advanced Nuclear Fuels GmbH (ANF). (author)

Modeling the Thermal Rocket Fuel Preparation Processes in the Launch Complex Fueling System

Directory of Open Access Journals (Sweden)

A. V. Zolin

2015-01-01

Full Text Available It is necessary to carry out fuel temperature preparation for space launch vehicles using hydrocarbon propellant components. A required temperature is reached with cooling or heating hydrocarbon fuel in ground facilities fuel storages. Fuel temperature preparing processes are among the most energy-intensive and lengthy processes that require the optimal technologies and regimes of cooling (heating fuel, which can be defined using the simulation of heat exchange processes for preparing the rocket fuel.The issues of research of different technologies and simulation of cooling processes of rocket fuel with liquid nitrogen are given in [1-10]. Diagrams of temperature preparation of hydrocarbon fuel, mathematical models and characteristics of cooling fuel with its direct contact with liquid nitrogen dispersed are considered, using the numerical solution of a system of heat transfer equations, in publications [3,9].Analytical models, allowing to determine the necessary flow rate and the mass of liquid nitrogen and the cooling (heating time fuel in specific conditions and requirements, are preferred for determining design and operational characteristics of the hydrocarbon fuel cooling system.A mathematical model of the temperature preparation processes is developed. Considered characteristics of these processes are based on the analytical solutions of the equations of heat transfer and allow to define operating parameters of temperature preparation of hydrocarbon fuel in the design and operation of the filling system of launch vehicles.The paper considers a technological system to fill the launch vehicles providing the temperature preparation of hydrocarbon gases at the launch site. In this system cooling the fuel in the storage tank before filling the launch vehicle is provided by hydrocarbon fuel bubbling with liquid nitrogen. Hydrocarbon fuel is heated with a pumping station, which provides fuel circulation through the heat exchanger-heater, with

DUPIC nuclear fuel manufacturing and process technology development

International Nuclear Information System (INIS)

Yang, Myung Seung; Park, J. J.; Lee, J. W.

2000-05-01

In this study, DUPIC fuel fabrication technology and the active fuel laboratory were developed for the study of spent nuclear fuel. A new nuclear fuel using highly radioactive nuclear materials can be studied at the active fuel laboratory. Detailed DUPIC fuel fabrication process flow was developed considering the manufacturing flow, quality control process and material accountability. The equipment layout of about twenty DUPIC equipment at IMEF M6 hot cell was established for the minimization of the contamination during DUPIC processes. The characteristics of the SIMFUEL powder and pellets was studied in terms of milling conditions. The characteristics of DUPIC powder and pellet was studied by using 1 kg of spent PWR fuel at PIEF nr.9405 hot cell. The results were used as reference process conditions for following DUPIC fuel fabrication at IMEF M6. Based on the reference fabrication process conditions, the main DUPIC pellet fabrication campaign has been started at IMEF M6 using 2 kg of spent PWR fuel since 2000 January. As of March 2000, about thirty DUPIC pellets were successfully fabricated

DUPIC nuclear fuel manufacturing and process technology development

Energy Technology Data Exchange (ETDEWEB)

Yang, Myung Seung; Park, J. J.; Lee, J. W. [and others

2000-05-01

In this study, DUPIC fuel fabrication technology and the active fuel laboratory were developed for the study of spent nuclear fuel. A new nuclear fuel using highly radioactive nuclear materials can be studied at the active fuel laboratory. Detailed DUPIC fuel fabrication process flow was developed considering the manufacturing flow, quality control process and material accountability. The equipment layout of about twenty DUPIC equipment at IMEF M6 hot cell was established for the minimization of the contamination during DUPIC processes. The characteristics of the SIMFUEL powder and pellets was studied in terms of milling conditions. The characteristics of DUPIC powder and pellet was studied by using 1 kg of spent PWR fuel at PIEF nr.9405 hot cell. The results were used as reference process conditions for following DUPIC fuel fabrication at IMEF M6. Based on the reference fabrication process conditions, the main DUPIC pellet fabrication campaign has been started at IMEF M6 using 2 kg of spent PWR fuel since 2000 January. As of March 2000, about thirty DUPIC pellets were successfully fabricated.

The advanced neutron source three-element-core fuel grading

International Nuclear Information System (INIS)

Gehin, J.C.

1995-01-01

The proposed Advanced Neutron Source (ANS) pre-conceptual design consists of a two-element 330 MW f nuclear reactor fueled with highly-enriched uranium and is cooled, moderated, and reflected with heavy water. Recently, the ANS design has been changed to a three-element configuration in order to permit a reduction of the enrichment, if required, while maintaining or improving the thermal-hydraulic margins. The core consists of three annular fuel elements composed of involute-shaped fuel plates. Each fuel plate has a thickness of 1.27 mm and consists of a fuel meat region Of U 3 Si 2 -Al (50% enriched in one case that was proposed) and an aluminum filler region between aluminum cladding. The individual plates are separated by a 1.27 mm coolant channel. The three element core has a fuel loading of 31 kg of 235 U which is sufficient for a 17-day fuel cycle. The goal in obtaining a new fuel grading is to maximize important temperature margins. The limits imposed axe: (1) Limit the temperature drop over the cladding oxide layer to less than 119 degrees C to avoid oxide spallation. (2) Limit the fuel centerline temperature to less than 400 degrees C to avoid fuel damage. (3) Limit the cladding wall temperature to less than the coolant. incipient-boiling temperature to avoid coolant boiling. Other thermal hydraulic conditions, such as critical heat flux, are also considered

Advanced teleoperation in nuclear applications: consolidated fuel reprocessing program

International Nuclear Information System (INIS)

Hamel, W.R.; Feldman, M.J.; Martin, H.L.

1984-01-01

A new generation of integrated remote maintenance systems is being developed to meet the needs of future nuclear fuel reprocessing at the Oak Ridge National Laboratory. Development activities cover all aspects of an advanced teleoperated maintenance system with particular emphasis on a new force-reflecting servomanipulator concept. The new manipulator, called the advanced servomanipulator, is microprocessor controlled and is designed to achieve force-reflection performance near that of mechanical master/slave manipulators. The advanced servomanipulator uses a gear-drive transmission which permits modularization for remote maintainability (by other advanced servomanipulators) and increases reliability. Human factors analysis has been used to develop an improved man/machine interface concept based upon colographic displays and menu-driven touch screens. Initial test and evaluation of two advanced servomanipulator slave arms and several other development components have begun. 9 references, 5 figures

Performance of advanced oxide fuel pins in EBR-II

International Nuclear Information System (INIS)

Lawrence, L.A.; Jensen, S.M.; Hales, J.W.; Karnesky, R.A.; Makenas, B.J.

1986-05-01

The effects of design and operating parameters on mixed-oxide fuel pin irradiation performance were established for the Hanford Engineering Development Laboratory (HEDL) advanced oxide EBR-II test series. Fourteen fuel pins breached in-reactor with reference 316 SS cladding. Seven of the breaches are attributed to FCMI. Of the remaining seven breached pins, three are attributed to local cladding over-temperatures similar to the breach mechanism for the reference oxide pins irradiated in EBR-II. FCCI was found to be a contributing factor in two high burnup, i.e., 11.7 at. % breaches. The remaining two breaches were attributed to mechanical interaction of UO 2 fuel and fission products accumulated in the lower cladding insulator gap, and a loss of cladding ductility possibly due to liquid metal embrittlement. Fuel smear density appears to have the most significant impact on lifetime. Quantitative evaluations of cladding diameter increases attributed to FCMI, established fuel smear density, burnup, and cladding thickness-to-diameter ratio as the major parameters influencing the extent of cladding strain

Influences of in-fuel physical-chemical processes on serviceability of energy reactor fuel elements

Energy Technology Data Exchange (ETDEWEB)

Bibilashvili, Yu K; Nekrasova, G A; Sukhanov, G I

1989-01-01

In-fuel physico-chemical processes and their effect on stress corrosion cracking of fuel element zirconium cladding are considered in the review. The mechanism of fission product release from the fuel is studied and the negative role of primarily iodine on the cladding corrosion process is demonstrated. Directions for improving the fuel element claddings and fuel to increase the fuel element serviceability are specified.

Influences of in-fuel physical-chemical processes on serviceability of energy reactor fuel elements

International Nuclear Information System (INIS)

Bibilashvili, Yu.K.; Nekrasova, G.A.; Sukhanov, G.I.

1989-01-01

In-fuel physico-chemical processes and their effect on stress corrosion cracking of fuel element zirconium cladding are considered in the review. The mechanism of fission product release from the fuel is studied and the negative role of primarily iodine on the cladding corrosion process is demonstrated. Directions for improving the fuel element claddings and fuel to increase the fuel element serviceability are specified

Application of the Advanced Distillation Curve Method to Fuels for Advanced Combustion Engine Gasolines

KAUST Repository

Burger, Jessica L.

2015-07-16

© This article not subject to U.S. Copyright. Published 2015 by the American Chemical Society. Incremental but fundamental changes are currently being made to fuel composition and combustion strategies to diversify energy feedstocks, decrease pollution, and increase engine efficiency. The increase in parameter space (by having many variables in play simultaneously) makes it difficult at best to propose strategic changes to engine and fuel design by use of conventional build-and-test methodology. To make changes in the most time- and cost-effective manner, it is imperative that new computational tools and surrogate fuels are developed. Currently, sets of fuels are being characterized by industry groups, such as the Coordinating Research Council (CRC) and other entities, so that researchers in different laboratories have access to fuels with consistent properties. In this work, six gasolines (FACE A, C, F, G, I, and J) are characterized by the advanced distillation curve (ADC) method to determine the composition and enthalpy of combustion in various distillate volume fractions. Tracking the composition and enthalpy of distillate fractions provides valuable information for determining structure property relationships, and moreover, it provides the basis for the development of equations of state that can describe the thermodynamic properties of these complex mixtures and lead to development of surrogate fuels composed of major hydrocarbon classes found in target fuels.

Advances in Metallic Fuels for High Burnup and Actinide Transmutation

Energy Technology Data Exchange (ETDEWEB)

Hayes, S. L.; Harp, J. M.; Chichester, H. J. M.; Fielding, R. S.; Mariani, R. D.; Carmack, W. J.

2016-10-01

Research and development activities on metallic fuels in the US are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is a desire to demonstrate a multifold increase in burnup potential. A number of metallic fuel design innovations are under investigation with a view toward significantly increasing the burnup potential of metallic fuels, since higher discharge burnups equate to lower potential actinide losses during recycle. Promising innovations under investigation include: 1) lowering the fuel smeared density in order to accommodate the additional swelling expected as burnups increase, 2) utilizing an annular fuel geometry for better geometrical stability at low smeared densities, as well as the potential to eliminate the need for a sodium bond, and 3) minor alloy additions to immobilize lanthanide fission products inside the metallic fuel matrix and prevent their transport to the cladding resulting in fuel-cladding chemical interaction. This paper presents results from these efforts to advance metallic fuel technology in support of high burnup and actinide transmutation objectives. Highlights include examples of fabrication of low smeared density annular metallic fuels, experiments to identify alloy additions effective in immobilizing lanthanide fission products, and early postirradiation examinations of annular metallic fuels having low smeared densities and palladium additions for fission product immobilization.

Enhancement of MARS with an Advanced Fuel Model by Coupling FRAPTRAN

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyong Chol; Lee, Young Jin; Han, Sam Hee [KAERI, Daejeon (Korea, Republic of)

2016-05-15

FRAPTRAN calculates heat conduction, heat transfer from cladding to coolant, elastic-plastic fuel and cladding deformation, cladding oxidation, fission gas release, and fuel rod gas pressure. FRAPTRAN is used for analyzing the fuel response under postulated accidents such as reactivity-initiated accidents (RIAs) and loss-of-coolant accidents (LOCAs), and also for analyzing and interpreting experimental results. Burnup dependent variables such as fuel densification and swelling, and cladding creep and irradiation growth may be considered by incorporating FRAPCON steady state depletion calculation results as the initial conditions. FRAPTRAN-DLL has been successfully verified and the coupled calculations have shown to provide reasonable results. An EOC core loaded with irradiated fuels was analyzed with the integrated code system. The coupled code system has demonstrated its applicability to variety of applications such as assessing the effects of fuel thermal conductivity degradation with burnup. MARS has been enhanced with the advanced fuel model of FRAPTRAN so that users can use the fuel rod performance evaluation capability in the transient analyses.

Advancements in the behavioral modeling of fuel elements and related structures

International Nuclear Information System (INIS)

Billone, M.C.; Montgomery, R.O.; Rashid, Y.R.; Head, J.L.

1989-01-01

An important aspect of the design and analysis of nuclear reactors is the ability to predict the behavior of fuel elements in the adverse environment of a reactor system. By understanding the thermomechanical behavior of the different materials which constitute a nuclear fuel element, analysis and predictions can be made regarding the integrity and reliability of fuel element designs. The SMiRT conference series, through the division on fuel elements and the post-conference seminars on fuel element modeling, provided technical forums for the international participation in the exchange of knowledge concerning the thermomechanical modeling of fuel elements. This paper discusses the technical advances in the behavioral modeling of fuel elements presented at the SMiRT conference series since its inception in 1971. Progress in the areas of material properties and constitutive relationships, modeling methodologies, and integral modeling approaches was reviewed and is summarized in light of their impact on the thermomechanical modeling of nuclear fuel elements. 34 refs., 5 tabs

Transmutation Fuel Fabrication-Fiscal Year 2016

Energy Technology Data Exchange (ETDEWEB)

Fielding, Randall Sidney [Idaho National Lab. (INL), Idaho Falls, ID (United States); Grover, Blair Kenneth [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-12-01

ABSTRACT Nearly all of the metallic fuel that has been irradiated and characterized by the Advanced Fuel Campaign, and its earlier predecessors, has been arc cast. Arc casting is a very flexible method of casting lab scale quantities of materials. Although the method offers flexibility, it is an operator dependent process. Small changes in parameter space or alloy composition may affect how the material is cast. This report provides a historical insight in how the casting process has been modified over the history of the advanced fuels campaign as well as the physical parameters of the fuels cast in fiscal year 2016.